safety analysis report: non-proprietary version of nac-stc ... · the use of astm a276 304 ss for...

ANAC ffffl INTERNATIONAL

Atlanta Corporate Headquarters: 3950 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

June 2018

Revision 18B

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Enclosure 1 to ED20180063 Page 1 of 1

Enclosure 1

RAJ Responses for

NAC-STC SAR, Revision 18B

June 2018

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NACINTERNATIONAL

NON-PROPRIETARY RESPONSES TO THE

UNITED ST ATES NUCLEAR REGULATORY COMMISSION

REQUEST FOR ADDITIONAL INFORMATION

February 2018

NAC-STC Docket No: 71-9235

CoC No: 9235

FOR REVIEW OF THE CERTIFICATE OF COMPLIANCE NO. 9235, STC TRANSPORATION PACKAGE

(CoC NO. 9235 DOCKET NO. 71-9235)

June 2018

Page 1 of 18

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TABLE OF CONTENTS


Coe No: 9235

GENERAL INFORMATION ....................................................................................................................... 3

MATERIALS EVALUATION ..................................................................................................................... 4

THERMAL EVALUATION ........................................................................................................................ 8

ACCEPTANCE AND MAINTENANCE TESTS EVALUATION ........................................................... 10

Page 2 of 18

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NAC INTERNATIONAL RESPONSE TO


GENERAL INFORMATION


CoC No: 9235

1.1 Clarify callout to Note 2 in Drawing No. 423-870 Sheet 1 of I Revision 7 in Zone El.

Note 2 was deleted from the drawing but the callout "SEE NOTE 2" remains in the

revised drawing.

This information is needed to determine comp! iance with Title IO of the Code of Federal Regulations (10 CFR) 71.31 ( c ).

NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 1.1:

Reference to Note 2 in drawing zone El has been removed from the field of the drawing .

Page 3 of 18

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NAC INTERNATIONAL RESPONSE

TO REQUEST FOR ADDITIONAL INFORMATION

MATERIALS EVALUATION


CoC No: 9235

2.1 Clarify the operating temperatures for ASTM International (ASTM) A276 Type 304 SS

which was added to the bill of materials for Item #2 in Drawing No. 423-859, Revision 1, Sheet 1 of 1. Note that ASTM A276 304 SS is not included in American Society of

Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section II Part

D. Provide mechanical prope11ies over the expected range of operating temperatures if

the mechanical properties of this component are necessary at elevated temperatures.

This information is needed to determine compliance with IO CFR 71.31 (c) and 10 CFR 71.33(a)(5).


The use of ASTM A276 304 SS for the retaining rod washer material of the balsawood

impact limiter (Item 2, Drawing No. 423-859 Revision I) is consistent with retaining rod

washer material approved for the redwood impact I imiter (Item 9, Drawing No. 423-811

referenced on the package assembly via Drawing No. 423-900). Both washers are used

for the same function of attaching and retaining the respective impact limiter. Therefore,

the addition of the material option for the balsawood impact limiter configuration is consistent with the current approved system. As detailed in Chapter 9 of the STC SAR,

the STC transport system is designed in accordance with "ASME Boiler and Pressure Vessel Code," The American Society of Mechanical Engineers, 1989 and 1992, with

Addenda, for the STC cask directly loaded fuel configurations

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CoCNo: 9235

2.2 Update the specification for the forged stainless steel (SS). Table 2.3.2-2 identifies Type

304 SS under specification SA-336. The austenitic SS that were formerly listed under

SA-336 have been moved to SA-965 in 2008. SA-336 Type 304 SS is called out in

Drawing Nos. 423-802 sheet 1 of7 and 423-804 Sheet 1 of 3.

This information is needed to determine compliance with 10 CFR 71.31(c) and 10 CFR

71.33(a)(5).

NAC International Response to Acceptance and Maintenance Tests Evaluation RAJ 2.2:

As detailed in Chapter 9 of the STC SAR, the STC transport system is designed in

accordance with "ASME Boiler and Pressure Vessel Code," The American Society of

Mechanical Engineers, 1989 and 1992, with Addenda, for the STC cask directly loaded

fuel configurations. Therefore, the ASME SA-336 Type 304 SS material reference is

consistent with the design basis code year and no change is necessary. Note that later

versions of the ASME SA-336 material specification (post 2001 with the 2003 addenda)

provides direction to SA-965 for stainless steel forgings .

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CoC No: 9235

2.3 Revise Tables 2.3.2-2 and 2.3.2-3 footnotes to identify the correct sources of information

from the ASME B&PV code. Except for fatigue design tables and figures, the mechanical property tables in Section Ill Appendix I have been moved to Section II Part D in the 1992 addenda.

This information is needed to determine compliance with 10 CFR 71.31 ( c) and IO CFR 71.33(a)(5).


The footnotes of the tables in Section 2.3.2 will be revised to note the corresponding tables from ASME B&PV Code Section II, Part D as follows:

Table 2.3.2-1 Mechanical Properties of SA 240, Type 304 Stainless Steel

1 "ASME Boiler and Pressure Vessel Code," Section 11, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section Il, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section 11, Paii D, Table TM-I. s "ASME Boiler and Pressure Vessel Code," Section lIJ, Appendix I, Table 1-9.1. 6"ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-1. 1 "ASME Boiler and Pressure Vessel Code," Section 11, Part D, Table NF-1.


1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section 11, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section ll, Part D, Table 2A.

4 "ASME Boiler and Pressure Vessel Code," Section JI, Part D, Table TM-1.

s "ASME Boiler and Pressure Vessel Code," Section III, Appendix I, Table I-9.1.

6 "ASME Boiler and Pressure Vessel Code," Section II, Pa1i D, Table TE-1. 1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-1. s "Nuclear Materials Handbook," Volume 1, Design Data, Property Code 3304 .

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RAI 2.3 Continued


CoCNo: 9235

Table 2.3.2-3 Mechanical Propetiies of Type XM-19 Stainless Steel

1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U.

2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1.

J "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table 2A.

4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-1.

s "ASME Boiler and Pressure Vessel Code," Section III, Appendix I, Table I-9.1.

6 "ASME Boiler and Pressure Vessel Code," Section JI, Part D, Table TE-I.

7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-1.

s "Nuclear Materials Handbook," Volume 1, Design Data, Property Code 3304.

9 SA-182, FXM-19 stainless steel may be substituted for SA-240 XM-19 stainless steel provided that

the SA-182 material yield and ultimate strengths are equal to or greater than those of the SA-240

material. The SA-182 forging material and the SA-240 plate material are both XM-19 austenitic

stainless steels. Austenitic stainless steels do not experience a ductile-to-brittle transition for the

range of temperatures considered in this Safety Analysis Report. Therefore, fracture toughness is

not a concern .

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THERMAL EVALUATION


CoC No: 9235

3 .1 Provide justification and, if necessary, calculations to show how the thermal analysis in

Section 3.4.1.1.1.3, "Radial Neutron Shield," of the application remains bounding for

normal conditions of transport and hypothetical accident conditions calculated

temperatures considering the change to note 7 on Drawing No. 423-802, Sheet 1.

No justification and calculations were provided to show how the normal conditions of

transport and hypothetical accident conditions calculated temperatures remain bounding

considering the change, "Alternate pre-bonded thickness of 8mm to 10mm for 304 SS

and 6mm to 8mm for copper plates may be used.," to note 7 on Drawing No. 423 802,

Sheet 1.

This information is necessary to demonstrate compliance with 10 CFR 71.71 and 10 CFR

71.73 .


The effective thermal conductivity calculation for the radial neutron shield presented in SAR Section 3 .4.1.1.1.3 is based on the 8 mm stainless steel plate and 6 mm copper plate. The change in note 7 of Drawing No. 423-802, Revision 23 allows the thickness for the stainless steel plate and copper plate to be increased to 10 mm and 8 mm respectively. The thermal analysis for normal conditions of transport for the directly loaded STC configuration is performed using the thermal models described in SAR Section 3.4.1.1 and the maximum temperatures are provided in Section 3.4.2.1 (Tables 3.4-1, 3.4-2 and 3 .4-3). The increased effective thermal conductivity for the radial neutron shield provides a slightly more effective path for heat rejection from the cask shell to the ambient. This results in a slight reduction of the maximum temperatures presented in Section 3.4.2.1. Therefore, the temperatures presented in Section 3.4.2.1 remain bounding and no further analysis is required.

For the hypothetical accident, the thermal evaluation for the directly loaded STC

configuration uses the thermal model described in SAR Section 3 .5 .1.1.1 and the maximum

component temperatures are presented in Section 3 .5 .3 (Table 3 .5-1 ). The radial neutron

shield property used in the thermal model is discussed in Section 3.5.1.1.3. Note that the NS-4-FR in the neutron shield is conservatively considered to be present for the entire

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RAI 3.1 Continued


CoC No: 9235

30-minute fire condition to maximize the heat input to the model. At the end of the fire transient, the neutron shield is considered to be voided ofNS-4-FR, leaving only the

stainless steel/copper fins and stainless steel shell. The effective conductivity of the

neutron shield is calculated using the same method described in SAR Section 3.4.1.1.1.3 with the NS-4-FR material substituted by air. Note that the temperature results for steady state thermal analysis for the normal condition of transport are used as the initial condition for the fire transient analysis. As previously discussed, the temperatures for the normal condition will be reduced due to higher effective conductivity of the radial neutron shield. Due to the short duration of the fire (30 minutes), cask shells and contents peak temperatures occur after 30 minute fire condition. The heat rejection after

the fire is slightly enhanced. The drawing change to allow thicker stainless steel and copper plates will have an insignificant impact on the maximum component temperatures

for the hypothetical fire accident as presented in SAR Table 3 .5-1.

To further evaluate the effect of the drawing change allowing thicker stainless steel and coper plates in the radial neutron shield, a sensitivity analysis is performed using a threedimensional thermal model corresponding to the limiting configuration and heat load of the NAC-STC transport cask, i.e. the STC-HBU configuration as approved by CoC No . 9235, Revision No. 15, December 20, 2016. The three-dimensional model from the main body ofNAC Calculation No. 423-3000 Rev. 5 is used to perform a steady state analysis for the normal condition of transport, as well as a transient analysis for the fire accident. The model is a 180° half-symmetry full length model for the cask containing the loaded

basket. The governing Case H3 (see Calculation No. 423-3000) with the design basis heat load of 24 kW for the STC-HBU is used. For the sensitivity study, the only change

in the thermal model is the effective conductivity of the radial neutron shield, which were re-calculated using same method as described in SAR Section 3 .4.1.1.1.3 considering the increased thickness of stainless steel and copper plates. The analysis results indicate that the maximum fuel temperature decreased by 4°F (from 638°F to 634°F) for the normal condition of transport. The maximum fuel temperature for the fire accident remained approximately the same (decreased only 1 °F, from 698°F to 697°F). See Appendix AA ofNAC Calculation No. 423-3000 Rev. 6 for details of the steady state and transient analysis.

It is concluded that the change in note 7 on Drawing No. 423-802, Revision 23, Sheet 1

has an insignificant effect on the thermal performance of NAC-STC cask. The analysis results presented in SAR Section 3.4.1.2 for normal condition of transport and Section

3.5.3 for hypothetic fire accident remain bounding and no revision is required .

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ACCEPTANCE AND MAINTENANCE TESTS EVALUATION

NAC-STC Docket No: 7 l-9235

CoC No: 9235

8.1 Provide a justification for the position implied in the SAR that a visual inspection of the

radial neutron shield shell, described in Sections 8.2.6 and 8.2.7 of the application, would

provide the necessary information to detect deterioration of the heat transfer properties of

the package. In addition, provide specific acceptance criteria for visual inspections of an

NAC-STC packaging, include a demonstration that these criteria are adequate to verify

the thermal performance of the packaging.

Sections 8.2.6, "Post-fabrication Thermal Test," and 8.2.7, "Miscellaneous," of the

application do not provide a justification that the visual inspection of the radial neutron

shield shell will be able to detect conditions that might lead to the deterioration of the

heat transfer properties of the package, especially any deterioration of the fins that are

internal to the radial neutron shield shell and not visible by inspecting the radial neutron

shield shell. A specific acceptance criteria for the visual inspection, e.g. a more specific

acceptance criteria than described in Section 8.2. 7 of the application, " ... any crack, gauge

(assume "gouge" is meant), or gross deformation that could indicate damage of the heat

transfer fins ... ," was not provided. Also, it has not been demonstrated that visual

inspection is an effective method, with either general or specific acceptance criteria, in

order to verify that the thermal performance of the package has not deteriorated.

Table 3.8-4, "Maximum Component Temperatures - Normal Transport Conditions,

Maximum Decay Heat, Maximum Ambient Temperature, among Three Configurations -

the STC-HBU," of the SAR dated April 11, 2017 (ADAMS Accession No.

ML 171 l 6A075) shows that the radial neutron shield maximum temperature is 295°F,

which is close to the upper temperature limit for the radial neutron shield of 300°F as

indicated in the SAR (Section 3.8.3.2, "Safe Operating Range").

The heat transfer capabilities of the fabricated packaging will impact the overall thermal

performance of the assembled package and in order for components, like the radial

neutron shield, to remain below the allowable temperature limit, the package must be in a

condition commensurate with its design at all times; therefore, it is important for all

inspections to be able to determine if there could be any potential degradation in thermal

performance.

This information is necessary to demonstrate compliance with 10 CFR 71.71 .

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RAI 8-1 Continued


CoC No: 9235


In understanding the basis for selecting a visual inspection of the neutron shield shell as

the approach to determine continuing compliance with the casks thermal heat rejection

design, it is important to understand the mechanical attachment of the thermal fins to the

cask body and the neutron shield shell, as well as the nature of the materials used and the

extent of cask normal conditions with which cask operations are performed. The

robustness of the design would preclude any degradation of the weldments due to normal

cask operations. Any issues associated with the adequacy of the fabrication would be

discovered on the first article thermal inspection.

The fin is an explosive-bonded, bi-metallic ( copper & 304 stainless steel), component

which is attached to the cask body outer shell (304 stainless steel) using a full penetration

groove weld with 1/8" fillet reinforcement. The weld is PT examined per Section V,

Article 6, with acceptance criteria per Section III, NF-5350.

The neutron shield shell is then attached to the using one of two approved methods, full

penetration groove welds or a single full penetration double-bevel weld. Again, the weld

is penetrant examined per Section V, Article 6, with acceptance criteria per Section III,

NF-5350.

Thermal cycling would not impose significant stresses at the connections of these

members due to their being similar materials. Normal operations will not fail these

connections as there is no loading of the connections described. The instance of an impact capable of imposing loads resulting in deformation or weld failure would be clearly visible by means of a general visual inspection of the neutron shield .

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CoC No: 9235

8.2 Provide a description (in Section 8.2.6 of the application) of how the NAC-STC package

will be monitored during handling and transportation operations for the normal

conditions of transport described in 10 CFR 71.7l(c).

Section 8.2.6, "Post-fabrication Thermal Test," of the application does not demonstrate

how the package will be monitored during handling and transp011ation operations to

ensure compliance with each of the requirements in 10 CFR Part 71.71 (c). For example,

the heat condition (an ambient temperature of 100°F in still air with solar insolation)

could potentially be exceeded during transportation operations, yet without monitoring,

this would not be detected. In addition, monitoring of the 10 CFR 71.7l(b) initial

conditions may be necessary (see RAI 3, below).

This information is necessary to demonstrate compliance with 10 CFR 71.7l(c) .


There is no regulatory requirement to monitor the package to ensure compliance with 10

CFR Part 71.71 (c). NAC recognizes the importance of maintaining the package within the requirements of IO CFR Pai1 71. 71 ( c) during transport, however, it was not the intent

to state that a post-fabrication thermal test would be performed after exceeding the each

of the requirements of 10 CFR Part 71.7l(c). Therefore, NAC has revised the text in

Section 8.2.6 to read as follows:

"However, a post-fabrication thermal test shall be performed on an operational NAC

STC packaging if, during handling or transport operations, the packaging experiences an

adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield."

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CoC No: 9235

8.3 Address the 10 CFR 71.7l(b) initial conditions in Section .2.6 of the application and

describe action(s) taken if those conditions are exceeded during transport operations.

Section 8.2.6, "Post-fabrication Thermal Test," of the application describes that a thermal

test will be performed on an operational NAC-STC packaging if the conditions or tests of

10 CFR 71.71 are exceeded during transportation operations. Section 8.2.6 of the

application does not describe if any action is taken if the initial conditions in IO CFR

71.71 (b) are exceeded (see item 8-2, above to address monitoring of the 10 CFR 71.71 (b)

initial conditions during handling and transport operations), or describe why not taking any action(s) is justified.

This information is necessary to demonstrate compliance with 10 CFR 71.71(b) .


There is no regulatory requirement to monitor the package to ensure compliance with I 0

CFR Part 71.71 (c). NAC recognizes the importance of maintaining the package within

the requirements of 10 CFR Pai1 71.71 (c) during transport, however, it was not the intent

to state that a post-fabrication thermal test would be performed after exceeding the each

of the requirements of 10 CFR Part 71.7l(c). Therefore, NAC has revised the text in Section 8.2.6 to read as follows:

"However, a post-fabrication thermal test shall be performed on an operational NAC

STC packaging if, during handling or transport operations, the packaging experiences an

adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield."

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CoC No: 9235

8.4 Clarify the acceptance criteria for the outer closure lid and the outer bottom plate. The

applicant stated: "The outer closure lid and the outer bottom plate will be UT [ultrasonic

tested] to demonstrate their soundness as gamma shielding utilizing ASME Section V,

Article 23, acceptance criteria. Plate shall be accepted per NB-2530 and forgings will be

accepted per NB-2540. Each ASME specification provides requirements for testing

equipment, test method, acceptance criteria, material traceability, and supporting

documentation. SAR Section 8.1.5 .1 has been revised to reflect these requirements."

ASME B&PV Section V Article 23 is not referenced in Section Ill NB-2530 or NB-2540.

It is only referenced in NB-2585 with respect to the examination of bolts. Note that

NB 2532.1 references SA-578 which is included in Section V Article 23. NB-2540

references Section V Article 5. NB-2540 does not reference ASME B&PV Section V

Article 23 .

It appears that the appropriate examination and acceptance criteria would be as follows:

The outer closure lid will be UT examined in accordance with ASME B&PV NB-2542.1

and the acceptance standards of Section NB-2542.2. The outer bottom plate shall be

examined in accordance with NB-2532.1 with the acceptance standards of NB-2532.1 (b).

This information is needed to determine compliance with IO CFR 71.31 ( c) and 10 CFR

71.33(a)(5) and 10 CFR 71.5l(a)(2).


NAC concurs with the reviewer's observation and acknowledges the Section V, Article

23 and Article 5 disconnect. NAC has revised Section 8.1.5.1, paragraph 2, to read as

follows:

"A gamma scan test is not required for the cask inner closure lid, cask outer closure lid,

cask inner bottom forging, cask outer bottom forging, or cask outer bottom plate. These

components shall be ultrasonic tested to demonstrate their soundness as gamma shielding.

Ultrasonic testing shall be pe,formed per ASME B&PV NB-2542. 1 using the acceptance standards of Section NB-2542.2for forgings andASME B&PV NB-2532.1 using the acceptance standards of NB- 2532.l(b)for plates."

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Coe No: 9235

8.5 Clarify the acceptance criteria for UT of the package outer bottom plate and package

inner bottom forging welds to the package outer bottom forging. The applicant stated:

"The package outer bottom forging will be UT to demonstrate its soundness as a gamma

shield utilizing ASME B&PV Section V, Article 23, acceptance criteria. The forging will

be accepted per NB-2540." The ASME B&PV specification provides requirements for

testing equipment, test method, acceptance criteria, material traceability, and supporting

documentation. SAR Section 8.1.5 .1 has been revised to add the package outer bottom

forging to the UT requirements."

NB-2540 references ASME B&PV Section V, Article 5. NB-2540 does not reference ASME B&PV Section V, Article 23.

This information is needed to determine compliance with 10 CFR 71.31 ( c ) .

NAC International Response to Acceptance and Maintenance Tests Evaluation RA! 8.5:

As indicated in the response to Item 8.4, NAC concurs with the reviewer's observation

and acknowledges the Section V, Article 23 and Article 5 disconnect. NAC has revised 8.1.5.1, paragraph 2, to read as follows:

"A gamma scan test is not required for the cask inner closure lid, cask outer closure lid,

cask inner bottom forging, cask outer bottom forging, or cask outer bottom plate. These components shall be ultrasonic tested to demonstrate their soundness as gamma shielding.

Ultrasonic testing shall be performed per ASME B&PV NB-2542.1 using the acceptance standards o_[Section NB-2542.2for forgings and ASME B&PV NB-2532.1 using the acceptance standards of NB- 2532.1 (b) for plates."

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8.6 Provide additional clarification for the alternate lead pour procedures.


CoC No: 9235

Section 8.4.2.2 states that during the lead pour the bottom end of the filler-tube is kept

below the surface of the molten lead to preclude the formation of voids in the lead.

Clarify whether the same practice is necessary for the alternate procedure.

Section 8.4.3.2 states that the body weldment will be heated in a steady, uniform, and

controlled manner. Provide the allowable heating rates.

Section 8.4.3.2 states that the temperature of the entire body weldment is maintained

between 640°F (338°C) and 740°F (393°C) throughout the lead pour operations,

approximately. Provide clarification on what "approximately" is referring to in this

context.

Section 8.4.3.3 states that the cooldown rate is held steady, uniform and controlled

manner. Provide maximum cooldown rate and the maximum allowable temperature differential between the inner and outer shell.

This information is needed to determine compliance with IO CFR 71.31 ( c) and

71.33(a)(5).

NAC lnternational Response to Acceptance and Maintenance Tests Evaluation RA] 8.6:

The response to this RAJ was provided on March 6111 in submittal 18A .

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CoC No: 9235

8.7 Clarify acceptable testing methods for the transport impact limiter SS shell. Section

8.1.4.3 states that a leak test of the shell welds shall be performed to verify weld integrity.

Three acceptable test methods are included but only the first two are actually leak tests.

The third method listed is penetrant testing which is a non-destructive test method. If

penetrant testing is allowed in lieu of an actual leak test, provide the penetrant testing

acceptance criteria and explain why the penetrant testing is a suitable method in lieu of a

leak test.

This information is needed to determine compliance with IO CFR 71.31 ( c ).


It is intended that the impact limiter shells are not to be qualified as "containment or

pressure retaining vessels", but sealed against environmental effects during transport,

specifically ingress of moisture. The implementation of the leakage test allows a single

point evaluation of the shell integrity. Implementation of a penetrant test for pin-holes,

cracks and/or porosity will also provide a similar level of confidence in the shells

environmental integrity.

Section 8.1.4.3 has been revised to include addition details regarding penetrant testing

acceptance criteria as follows:

Liquid penetrant examined per ASME B&PV Section V, Article 6. Acceptance per

Section III, Article NF-5350 .

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Coe No: 9235

8.8 Provide justification for replacing Viton 0-rings at least once every 2 years, or revise Section 8.1.4.2 of the SAR to specify a replacement frequency of at least 1 year during

transport operations or prior to transport if they have been installed longer than 1 year.

The applicant removed reference to PTFE 0-rings in SAR Section 8.1.4.2. Due to this change, the third paragraph in the section now states: "Viton 0-rings shall be replaced at least once every two years during cask transport operations, or prior to transport if they have been installed longer than two years." The SAR Section 8.1.4.2 revision 18 from

March 2017 stated, "The Viton 0-rings shall be replaced at least annually during cask

transpo1i operations, or prior to transport if they have been installed longer than one year (i.e., for extended cask out of service periods)." Replacing elastomeric seals at an interval

not to exceed one year is consistent with Section 8.3.4.3, "Component Tests," of

NUREG-1617, "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel."

This information is needed to determine compliance with 10 CFR 71.43(f) and 10 CFR 71.51(a).


SAR section 8.1.4.2, third paragraph has been revised to read as follows:

"Those Vi ton 0-rings that provide the Containment Boundary seal shall be replaced annually during cask transport operations, or prior to transport of they have been installed longer than one year. Secondary Boundary (i.e., Non-Containment Boundary) Viton 0-rings shall be replaced at least once every two years during cask transport operations, or prior to transport if they have been installed longer than two years."

Page 18 of 18

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Enclosure 2 to ED20180063 Page I of2

Enclosure 2

List of Drawing Changes


June 2018

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Enclosure 2 to ED20180063 Page 2 of2

List of Drawing Changes, NAC-STC SAR, Revision 18B

Drawing 423-209, Revision 02

1. Zone B6, Revise dimension to "47.00 +.02/-.20 TYP", was "47.00". 2. Zone F2/3, Revise dimension to "076.0", was "076.00". 3. Zone F2, Revise dimension to "0124.00 +.20/-.02", was "0124.00". 4. Zone E5, Revise dimension to "044.00 +.20/-.02", was "044.00".

Drawing 423-210, Revision 02

1. Zone F2, Revise dimension to "076.0", was "076.00". 2. Zone F2, Revise dimension to "0124.00 +.20/-.02", was "0124.00". 3. Zone E5, Revise dimension to "044.00 +.20/-.02", was "044.00".

Drawing 423-870, Rev 8

1. Zone El, removed leader with text "See Note 2" .

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Enclosure 3

Supporting Calculations for


June 2018

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List of Calculations and Supporting Documents

I. Calculation 423-3000, Revision 6

Calculation withheld in its entirety per IO CFR 2.390 .

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Enclosure 4

Proposed Changes for Certificate of Compliance Revision 19


June 2018

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CoC Sections (revised)

Page 5 of 19 5.(a)(3) Drawings

(i) The cask is constructed and assembled in accordance with the following Nuclear Assurance Corporation (now NAC International) Drawing Nos.:

Page 6 of 19

423-800, sheets 1-3, Rev. 19P & 19NP 423-802, sheets 1-7, Rev. 25 423-803, sheets 1-2, Rev. 14 423-804, sheets 1-3, Rev. 12 423-805, sheets 1-2, Rev. 8 423-806, sheets 1-2, Rev. 13 423-807, sheets 1-3, Rev. 5

5.(a)(3) Drawings

423-811, sheets 1-2, Rev. 13 423-812, Rev. 7 423-900, Rev. 8 423-209, Rev. 2 423-210, Rev. 2 423-901, Rev. 3

(ii) For the directly loaded configuration, the basket is constructed and assembled in accordance with the following Nuclear Assurance Corporation (now NAC International) Drawing Nos.:

Page 7 of 19

423-870, Rev. 8 423-871, Rev. 5 423-872, Rev. 6 423-873, Rev. 2

5.(a)(3) Drawings (Continued)

423-874, Rev. 3 423-875, sheets 1-2, Rev. 11 423-878, sheets 1-2, Rev. 5 423-880, Rev. 2P & lNP

(v) The Balsa Impact Limiters are constructed and assembled in accordance with the following NAC International Drawing Nos.:

423-257, Rev. 3 423-258, Rev. 3

423-843, Rev. 6 423-859, Rev. 1

•

•

•


Enclosure 5

List of Changes


June 2018

• Enclosure 5 to ED20180063 Page 2 of2

List of Changes, NAC-STC SAR, Revision 18B

Chapter 1

• Page 1-v, modified List of Drawings to reflect drawing revision.

Chapter 2

• Pages 2.3.2-2 thru 2.3.2-4, updated Table Notes for Tables 2.3.2-1, 2.3.2-2 and 2.3.2-3.

Chapter 3

• No changes.

Chapter 4

• No changes.

Chapter 5

• No changes.

Chapter 6

• • No changes.

•

Chapter 7

• No changes.

Chapter 8

• Page 8.1-10, modified the third paragraph of Section 8.1.4.2. • Page 8.1-11, modified Item 3 of Section 8.1.4.3. • Page 8.1-12, modified the second paragraph of Section 8.1.5.1. • Pages 8.2-4 thru 8.2-5, modified Section 8.2.6. • Page 8.2-7, modified last row of Table 8.2-1.

Chapter 9

• No changes .

•

•

•

Enclosure 6 to ED20180063 Pagel of I

Enclosure 6

SAR Changed Pages and LOEP


June 2018

•

•

• ANAC fdFI I NTE RNATIO NAL

Atlanta Corporate Headquarters: 3950 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

June 2018

Revision l8B

•

•

•

NAC-STC SAR Docket No. 71-9235

June 2018 Revision l 8B

List of Effective Pages

Chapter 1 Page 2.4-1 ..................................... Revision 18

Page 2.4.1-1 .................................. Revision 18

Page 1-i thru I-iv .......................... Revision 18 Page 2.4.2-1 .................................. Revision 18

Page 1-v ...................................... Revision l 8B Page 2.4.3-1 .................................. Revision 18

Page I-vi thru I-ix ........................ Revision 18 Pages 2.4.4-1 thru 2.4.4-6 ............. Revision 18

Pages 1-1 thru 1-12 ....................... Revision 18 Pages 2.4.4-7 thru 2.4.4-8 ........... Revision l 8B

Pages 1.1-1 thru 1.1-3 ................... Revision 18 Page 2.4.4-9 .................................. Revision 18

Page 1.1-4 ................................... Revision l 8B Pages 2.4.4-10 ............................. Revision l 8B

Pages 1.1-5 thru 1.1-46 ................. Revision 18 Page 2.4.5-1 .................................. Revision 18

Pages 1 .2-1 thru 1.2-8 ................... Revision 18 Page 2.4.6-1 .................................. Revision 18

Page 1.2-9 ................................... Revision l 8B Pages2.5.l-l thru2.5.l-38 ........... Revision 18

Pages 1.2-10 thru 1.2-42 ............... Revision 18 Pages 2.5.2-1 thru 2.5.2-29 ........... Revision 18

Pages 1.2-43 thru 1.2-44 ............. Revision l 8B Pages 2.6-1 thru 2.6-2 ................... Revision 18

Pages 1.2-45 thru 1.2-49 ............... Revision 18 Pages 2.6.1-1 thru 2.6.1-7 ............. Revision 18

Page 1.3-1 ..................................... Revision 18 Pages 2.6.2-1 thru 2.6.2-8 ............. Revision 18


Page 2.6.4-1 .................................. Revision 18

Chapter 2 Pages 2.6.5-1 thru 2.6.5-2 ............. Revision 18

Page 2.6.6-1 .................................. Revision 18

Pages 2-i thru 2-lxviii .................... Revision 18 Page 2.6.7-1 .................................. Revision 18

Page 2-1 ........................................ Revision 18 Pages2.6.7.l-l thru 2.6.7.1-17 .... Revision 18

Pages2.1.1-l thru2.l.l-2 ............. Revision 18 Pages 2.6.7.2-1 thru 2.6.7.2-19 ..... Revision 18

Pages 2.1.1-3 thru 2.1.1-4 ........... Revision l 8B Pages 2.6. 7 .3-1 thru 2.6. 7 .3-11 ..... Revision 18

Pages 2.1.1-5 ................................. Revision 18 Pages 2.6.7.4-1 thru 2.6. 7.4-59 ..... Revision 18

Pages 2.1.2-1 thru 2.1.2-5 ............. Revision 18 Pages 2.6.7.5-1 thru 2.6.7.5-4 ....... Revision 18

Pages 2.1.3-1 thru 2.1.3-15 ........... Revision 18 Pages 2.6.7.5-5 thru 2.6.7.5-8 ..... Revision l 8B

Pages 2.2-1 thru 2.2-8 ................... Revision 18 Pages 2.6.7.5-9 thru 2.6.7.5-13 ..... Revision 18

Pages 2.3 .1-1 thru 2.3 .1-2 ............. Revision 18 Pages 2.6.7.6-1 thru 2.6.7.6-13 ..... Revision 18

Pages 2.3.2-1 ................................. Revision 18 Pages 2.6.7.7-1 thru 2.6.7.7-5 ....... Revision 18

Pages 2.3.2-2 thru 2.3.2-4 ........... Revision l 8B Page 2.6.8-1 .................................. Revision 18

Pages 2.3.2-5 ................................. Revision 18 Page 2.6.9-1 .................................. Revision 18

Pages 2.3.3-1 thru 2.3.3-2 ............. Revision 18 Page 2.6.10-1 ................................ Revision 18

Pages 2.3.4-1 thru 2.3.4-3 ............. Revision 18 Pages 2.6.10.1-1 thru

Pages 2.3.5-1 thru 2.3.5-2 ............. Revision 18 2.6.10.1-2 ................................ Revision 18

Pages 2.3 .6-1 thru 2.3 .6-5 ............. Revision 18 Pages 2.6.10.2-1 thru

Page 2.3.7-1 .................................. Revision 18 2.6.10.2-4 ................................ Revision 18

Page 2.3.8-1 .................................. Revision 18

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June2018 Revision 18B

List of Effective Pages (continued)

Pages 2.6.10.3-1 thru Pages 2.6.13.2-1 thru

2.6.10.3-7 ................................. Revision 18 2.6.13.2-7 ................................. Revision 18

Page 2.6.11-1 .............................. Revision 18B Pages 2.6.13.3-1 thru

Pages 2.6.11.1-1 thru 2.6.13.3-4 ................................. Revision 18

2.6.11.1-4 .............................. Revision 18B Pages 2.6.13.4-1 thru


2.6.11.2-11 ............................ Revision 18B Pages 2.6.13 .5-1 thru

Page 2.6.11.3-1 .............................. Revision 18 2.6.13 .5-2 ................................. Revision 18

Pages 2.6.12-1 thru Pages 2.6.13.6-1 thru

2.6.12-5 .................................... Revision 18 2.6.13 .6-2 ................................. Revision 18

Page 2.6.12.1-1 .............................. Revision 18 Pages 2.6.13.7-1 thru

Pages 2.6.12.2-1 thru 2.6.13. 7-2 ................................. Revision 18

2.6.12.2-5 ................................. Revision 18 Page 2.6.13.8-1 .............................. Revision 18

Pages 2.6.12.3-1 thru Page 2.6.13.9-1 .............................. Revision 18

2.6.12.3-7 ................................. Revision 18 Page 2.6.13.10-1 ............................ Revision 18


2.6.12.4-3 ................................. Rev is ion 18 2.6.13 .11-3 ............................... Revision 18


2 .6.12.5-3 ................................. Revision 18 2.6.13 .12-2 ............................... Revision 18

Pages 2.6.12.6-1 thru Pages 2.6.14-1 thru

2.6.12.6-2 ................................. Revision 18 2. 6 .14-8 .................................... Revision 1 8


2.6. l 2.7-22 ............................... Revision 18 2 .6 .14 .1-2 ................................. Revision 18


2 .6 .12.8-2 ................................. Revision 18 2.6.14.2-16 ............................... Revision 18


2.6.12.9-11 ............................... Revision 18 2.6.14.3-3 ................................. Revision 18

Page 2.6.12.10-1 ............................ Revision 18 Pages 2.6.14.4-1 thru

Page 2.6.12.11-1 ............................ Revision 18 2.6.14.4-4 ................................. Revision I 8



2.6. l 2.13-4 ............................... Revision 18 Page 2.6.14.6-1 .............................. Revision 18


2.6.13-3 .................................... Revision 18 2.6.14. 7-14 ............................... Revision 18


2.6.13.1-2 ................................. Revision 18 2 .6.14.8-6 ................................. Revision 18

Page 2 .6.14.9-1 .............................. Revision 18

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June 2018 Revision 18B

List of Effective Pages ( continued)

Page 2.6.14.10-1 ........................... Revision 18 Pages 2.6. 16.6-1 thru

Pages 2.6.14.11-1 thru 2.6.16.6-3 ................................ Revision 18

2.6.14.11-5 .............................. Revision 18 Pages 2.6.16.7-1 thru

Pages 2.6.14.12-1 thru 2.6.16.7-12 .............................. Revision 18


Page 2.6.15-1 ................................ Revision 18 2.6.16.8-7 ................................ Revision 18

Pages 2.6.15.1-1 thru Page 2.6.16.9-1 ............................. Revision 18

2.6.15.1-2 ................................ Revision 18 Page 2.6.16. 10-1 ........................... Revision 18

Pages 2.6.15 .2-1 thru Pages 2.6.16. 11-1 thru

2.6.15 .2-7 ................................ Revision 18 2.6.16.11-4 .............................. Revision 18

Pages 2.6.15.3-1 thru Pages 2.6.16.12- 1 thru

2.6.15.3-4 ................................ Revision 18 2.6.16.12-2 .............................. Revision 18


2.6.15 .4-4 ................................ Revision 1 8 2.6.16.13-2 .............................. Revision 18

Page 2.6.15.5-1 ............................. Revision 18 Page 2.6.16.14-1 ........................... Revision 18


2.6. 15 .6-3 ................................ Revision 18 2.6.17-13 ................................. Revision 18

Page2.6.15.7-1 ............................. Revision 18 Pages 2.6.18-1 thru

Page2.6.15.8-1 ............................. Revision 18 2.6.18-6 ................................... Revision 18

Page2.6.15.9-1 ............................. Revision 18 Pages 2.6.19-1 thru

Page 2.6.15.10-1 ........................... Revision 18 2.6.19-23 ................................. Revision 18

Pages2.6.15.11-l thru Pages 2.6.20-1 thru

2.6.15.11-3 .............................. Revision 18 2.6.20-20 ................................. Revision 18


2.6.15 .12-2 .............................. Revision 18 2.6.21.-2 .................................. Revision 18

Pages 2.6.16-1 thru Pages2.7-1 thru2.7-2 ................... Revision 18

2.6.16-6 ................................... Revision 18 Page 2.7.1-1 thru 2.7.1-2 ............... Revision 18


2.6.16.1-2 ................................ Revision 18 2.7.1.1-15 ................................ Revision 18


2.6. 16.2-11 .............................. Revision 18 2.7.1.2-15 ................................ Revision 18


2.6.16.3-3 ................................ Revision 18 2.7. 1 .3-9 .................................. Revision 18


2.6.16.4-3 ................................ Revision 18 2.7.1 .4-11 ................................ Revision 18


2.6.15.5-3 ................................ Revision 18 2.7.1.5-3 .................................. Revision 18

3 of7


June 2018 Revision I 8B


Pages 2.7.1.6-1 thru

2.7. l .6-16 ................................. Revision 18

Page 2.7.2-1 ................................... Revision 18


2. 7 .2.1-5 ................................... Revision 18


2. 7 .2 .2-9 ................................... Revision 18


2. 7 .2.3-6 ................................... Revision 18


2.7 .2.4-7 ................................... Revision 18

Page 2.7.2.5-1 ................................ Revision 18

Page 2.7.2.6-1 ................................ Revision 18

Page 2.7.3.1-1 ................................ Revision 18

Pages 2.7.3.2-1 thru 2.7.3.2-5 ........ Revision 18



Page 2.7.3.5-1 ................................ Revision 18

Page 2.7.3.6-1 ................................ Revision 18

Page 2. 7.4-1 ................................... Revision 18

Page 2. 7 .5-1 ................................... Revision 18

Page 2. 7 .6-1 ................................... Revision 18

Pages 2.7.7-1 thru 2.7.7-4 .............. Revision 18


Pages 2.7.8. 1-1 thru 2.7.8.1-43 ...... Revision I 8


Pages 2.7.8.3-1 thru 2.7.8.3-13 ...... Revision 18

Pages 2.7.8.4-1 thru 2.7.8.4-1 O ...... Revision 18

Page 2.7.8.5-1 ................................ Revision 18

Pages 2.7.9-1 thru 2.7.9-40 ............ Revision 18

Pages 2.7.10-1 thru 2.7.10-12 ........ Revision 18

Pages2.7.11-1 thru2.7.ll-16 ........ Revision 18


Pages 2.7.13-1 thru 2.7.13-4 .......... Revision 18


2. 7 .13 .2-2 ................................. Revision 18


2.7.13.3-4 ................................. Revision 18


2. 7 .13 .4-8 ................................. Revision 18


2.7.13.5-2 ................................. Revision 18

Pages 2.7.14-1 thru

2.7.14-13 .................................. Revision 18

Pages 2.7.15-1 thru

2. 7 .15-16 .................................. Revision 18

Page 2.8-1 ...................................... Revision 18

Pages 2.9-1 thru 2.9-11 .................. Revision 18




Pages 2.10.4-1 thru

2. I 0.4-288 ................................ Revision 18


Pages 2.10.6-1 thru 2.10.6.-36 ....... Revision 18

13 drawings in Sections

2. I 0.6.6 and 2.10.6.7

Pages 2.10.6-37 thru

2. l O .6-8 8 .................................. Revision 18





2.10.10-11 ................................ Revision 18


2.10.11-8 .................................. Revision 18


•

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2.7.13.1-18 ............................... Revision 18 2.10.12-31 ................................ Revision 18 •

4 of7

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4 drawings in Section 2.10.12 Chapter 3

Pages 2.11.1-1 thru 2.11.1-2 ......... Revision 18 Page 3-i ....................................... Revision 18B

Pages 2.11.2-1 thru 2.11.2-2 ......... Revision 18 Pages 3-ii thru 3-iii ........................ Revision 18

Page 2.11.3-1 ................................ Revision 18 Pages 3-iv thru 3-v ...................... Revision 18B

Page 2.11.4-1 ................................ Revision 18 Page 3-vi ....................................... Revision 18

Page 2.11.5-1 ................................ Revision 18 Page 3-vii .................................... Revision 18B

Pages 2.11.6-1 thru 2.11.6-6 ......... Revision 18 Page 3-viii ..................................... Revision 18

Page 2.11.6.12-1 thru Pages 3.1-1 thru 3.1-12 ................. Revision 18

2.ll.6.12-62 ............................ Revision 18 Pages 3 .2-1 thru 3 .2-14 ................. Revision 18

Pages 2.11.6.13-1 thru Pages 3.3-1 thru 3.3-6 ................... Revision 18

2.11.6.13-35 ............................ Revision 18 Pages 3 .4-1 thru 3 .4-44 ................. Revision 18

Pages 2.11.6.14-1 thru Pages 3 .4-45 thru 3 .4-86 ............. Revision 18B

2.11.6.14-10 ............................ Revision 18 Pages 3.5-1 thru 3.5-16 ................. Revision 18

Page 2.11.6.15-1 ........................... Revision 18 Page 3 .6-1 ..................................... Revision 18

Pages 2.11. 7-1 thru 2.11. 7-8 ......... Revision 18 Pages 3.6.1-1 thru 3.6.1-4 ............. Revision 18

Pages 2.11.7.8-1 thru Pages 3 .6.2-1 thru 3 .6.2-3 ............. Revision 18

2.11. 7 .8-34 .............................. Revision 18 Pages 3.6.3-1 thru 3.6.3-3 ............. Revision 18

Pages 2.11.7.9-1 thru Pages 3 .6.4-1 thru 3 .6.4-24 ........... Revision 18

2.ll.7.9-14 .............................. Revision 18 Pages 3.6.5-1 thru 3.6.5-3 ............. Revision 18

Pages 2.11.7 .10-1 thru Page 3.7-1 ..................................... Revision 18

2.11.7.10-5 .............................. Revision 18 Pages 3.7.1-1 thru 3.7.1-3 ............. Revision 18

Page 2.11.8-1 ................................ Revision 18 Pages 3.7.2-1 thru 3.7.2-2 ............. Revision 18

Pages 2.11.9-1 thru 2.11.9-10 ....... Revision 18 Pages 3.7.3-1 thru 3.7.3-2 ............. Revision 18

Page2.12.l-l ................................ Revision 18 Pages 3.7.4-1 thru 3.7.4-9 ............. Revision 18

Pages 2.12.2-1 thru 2.12.2-2 ......... Revision 18 Page 3.7.5-1 thru 3.7.5-2 ............... Revision 18

Page 2.12.3-1 ................................ Revision 18 Page 3.8-1 ..................................... Revision 18

Page 2.12.4-1 ................................ Revision 1 8 Pages 3.8.1-1 thru 3.8.1-4 ............. Revision 18


Page 2.12.6-1 thru 2.12.6-29 ......... Revision 18 Pages 3.8.3-1 thru 3.8.3-3 ............. Revision 18

Page 2.13.1-1 ................................ Revision 18 Pages 3.8.4-1 thru 3.8.4-17 ........... Revision 18

Pages 2.13.2-1 thru 2.13.2-2 ......... Revision 18 Pages 3.8.5-1 thru 3.8.5-2 ............. Revision 18

Page 2.13.3-1 ................................ Revision 18 Page 3.8.6-1 .................................. Revision 18

Page 2.13.4-1 ................................ Revision 18

Page 2.13 .5-1 ................................ Revision 18

Pages 2.13.6-1 thru 2.13.6-62 ....... Revision 18

5 of7




Chapter 4 Pages 5.7.6-1 thru 5.7.6-22 ............ Revision 18

Page 5 .8-1 ...................................... Revision 18

Page 4-i thru 4-iii ........................... Revision 18 Pages 5.8.1-1 thru 5.8.1-9 .............. Revision 18

Pages 4.1-1 thru 4.1-10 .................. Revision 18 Pages 5.8.2-1 thru 5.8.2-7 .............. Revision 18


Pages 4.3-1 thru 4.3-4 .................... Revision 18 Pages 5.8.4-1 thru 5.8.4-3 .............. Revision 18

Page 4.4-1 ...................................... Revision 18 Pages 5.8.5-1 thru 5.8.5-4 .............. Revision 18


Page 4.5-14 ................................. Revision l8B Pages 5.8.7-1 thru 5.8.7-5 .............. Revision 18

Pages 4.5-15 thru 4.5-35 ................ Revision 18 Pages 5.8.8-1 thru 5.8.8-5 .............. Revision 18

Pages 4.6-1 thru 4.6-2 .................... Revision 18 Pages 5.8.9-1 thru 5.8.9-26 ............ Revision 18

Pages 4.7-1 thru 4.7-3 .................... Revision 18

Chapter 6

Chapter 5

Pages 6-i thru 6-ix ......................... Revision 18

Page 5-i thru 5-xix ......................... Revision 18 Pages 6.1-1 thru 6.1-6 .................... Revision 18

Pages 5-1 thru 5-4 .......................... Revision 18 Pages 6.2-1 thru 6.2-11 .................. Revision 18

Pages 5 .1-1 thru 5.1-30 .................. Revision 18 Pages 6.3-1 thru 6.3-10 .................. Revision 18

Pages 5 .2-1 thru 5 .2-40 .................. Revision 18 Pages 6.4-1 thru 6.4-2 .................... Revision 18

Pages 5.3-1 thru 5.3-33 .................. Revision 18 Page 6.4.1-1 ................................... Revision 18

Pages 5 .4-1 thru 5 .4-3 .................... Revision 18 Pages 6.4.2-1 thru 6.4.2-11.. .......... Revision 18

Page 5 .4-4 ................................... Revision l 8B Pages 6.4.3-1 thru 6.4.3-29 ............ Revision 18

Pages 5.4-5 thru 5.4-41 .................. Revision 18 Pages 6.4.4-1 thru 6.4.4-30 ............ Revision 18

Pages 5.5-1 thru 5.5-61.. ................ Revision 18 Pages 6.5-1 thru 6.5-2 .................... Revision 18

Page 5 .6-1 ...................................... Revision 18 Pages 6.5.1-1 thru 6.5.1-21.. .......... Revision 18

Pages 5 .6.1-1 thru 5 .6.1-9 .............. Revision 18 Pages 6.5.2-1 thru 6.5.2-20 ............ Revision 18

Pages 5.6.2-1 thru 5.6.2-20 ............ Revision 18 Pages 6.6-1 thru 6.6-2 .................... Revision 18

Pages 5.6.3-1 thru 5.6.3-13 ............ Revision 18 Pages 6.7-1 thru 6.7-333 ................ Revision 18

Pages 5.6.4-1 thru 5.6.4-34 ............ Revision 18 Page 6.8-1 ...................................... Revision 18

Page 5.6.5-1 ................................... Revision 18 Pages 6.8.1-1 thru 6.8.1-6 .............. Revision 18

Pages 5.6.6-1 thru 5.6.6-57 ............ Revision 18 Pages 6.8.2-1 thru 6.8.2-2 .............. Revision 18

Page 5.7-1 ...................................... Revision 18 Pages 6.8.3-1 thru 6.8.3-20 ............ Revision 18

Pages 5.7.1-1 thru 5.7.1-5 .............. Revision 18 Pages 6.8.4-1 thru 6.8.4-34 ............ Revision 18


Pages 5.7.3-1 thru 5.7.3-10 ............ Revision 18 Page 6.8.6-1 ................................... Revision 18

Pages 5.7.4-1 thru 5.7.4-14 ............ Revision 18 Pages 6.8.7-1 thru 6.8.7-27 ............ Revision 18

Page 5. 7 .5-1 ................................... Revision 18 Page 6.9-1 ...................................... Revision 18

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Page 6.9 .1-1 .................................. Revision 18

Page 6.9.2-1 .................................. Revision 18

Chapter 7

Pages 7-i thru 7-ii .......................... Revision 18

Page 7-1 thru 7-3 ........................... Revision 18

Pages 7 .1-1 thru 7.1-18 ................. Revision 18

Pages 7 .2-1 thru 7 .2-5 ................... Revision 18

Pages 7.3-1 thru 7.3-10 ................. Revision 18


Page 7 .5-1 ..................................... Revision 18


Chapter 8

Page 8-i thru 8-ii ......................... Revision 18B

Page 8-1 ........................................ Revision 18

Pages 8.1-1 thru 8.1-5 ................... Revision 18

Pages 8.1-6 thru 8.1-7 ................. Revision 18B


Pages 8.1-10 thru 8.1-25 ............. Revision 18B

Pages 8.1-26 thru 8.1-38 ............... Revision 18



Page 8.3-1 ..................................... Revision 18


Page 8.4-4 ..................................... Revision 18

Pages 8.4-5 thru 8.4-14 ............... Revision 18B

Chapter 9

Page 9-i ......................................... Revision 18

Pages 9-1 thru 9-13 ....................... Revision 18

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NAC-STC SAR

Docket No. 71-9235, Revisions 17C and 18A

June2018

Revision 18B

List of Drawings

Revision Drawing Number No. Title

423-800, sheets 1-3 Rev 19pO) Cask Assembly - NAC-STC Cask

423-800, sheets 1-3 Rev 19NP0l Cask Assembly- NAC-STC Cask

423-802, sheets 1-7 Rev 25 Cask Body - NAC-STC Cask

423-803, sheets 1-2 Rev 14 Lid Assembly - Inner, NAC-STC Cask

423-804, sheets 1-3 Rev 12 Details - Inner Lid, NAC-STC Cask

423-805, sheets 1-2 Rev 8 Lid Assembly - Outer, NAC-STC Cask

423-806, sheets 1-2 Rev 13 Port Coverplate Assy - Inner Lid, NAC-STC Cask

423-807, sheets 1-3 Rev 5 Assembly, Port Cover, NAC-STC Cask

423-209 Rev2 Impact Limiter Assy - Upper, NAC-STC Cask

423-210 Rev2 Impact Limiter Assy - Lower, NAC-STC Cask

423-257 Rev 3 Balsa Impact Limiter, Upper, NAC-STC Cask

423-258 Rev 3 Balsa Impact Limiter, Lower, NAC-STC Cask

423-811, sheets 1-2 Rev 13 Details - NAC-STC Cask

423-812 Rev 7 Nameplates -NAC-STC Cask

423-843 Rev 6 Transport Assembly, Balsa Impact Limiters, NAC-STC

423-859 Rev 1 Attachment Hardware, Balsa Limiters, NAC-STC

423-870 Rev 8 Fuel Basket Assembly, PWR, 26 Element, NAC-STC Cask

423-871 Rev 5 Bottom Weldment, Fuel Basket, PWR, 26 Element, NAC-STC Cask

423-872 Rev 6 Top Weldment, Fuel Basket, PWR, 26 Element, NAC-STC Cask

423-873 Rev 2 Support Disk and Misc. Basket Details, PWR, 26 Element, NAC-STC Cask

423-874 Rev 3 Heat Transfer Disk, Fuel Basket, PWR, 26 Element, NAC-STC Cask

423-875, sheets 1-2 Rev 11 Tube, NAC-STC Cask

423-878, sheets 1-2 Rev 5 Alternate Tube Assembly, NAC-STC Cask

423-880 Rev 2p(I) Shielded Thermal Shunt Assembly, NAC-STC Cask

423-880 Rev lNP(I) Shielded Thermal Shunt Assembly, NAC-STC Cask

423-900 Rev 8 Package Assembly Transportation, NAC-STC Cask

423-901, sheets 1-2 Rev 3 Transportation Package Concept, NAC-STC Cask

455-800, sheets 1-2 Rev 2 Assembly, Transport Cask, MPC-Yankee

(I) Proprietary and Non-proprietary drawing versions are only included in their respective SAR versions .

1-v

NAC-STC SAR

Docket No. 71-9235

Drawing Number 455-801, sheets 1-2 455-820, sheets 1-2

455-870

455-871, sheets 1-2

455-871, sheets 1-3

455-872, sheets 1-2

455-872, sheets 1-2

455-873

455-881, sheets 1-3

455-887, sheets 1-3

455-888, sheets 1-2

455-891, sheets 1-2

455-891, sheets 1-3

455-892, sheets 1-2

455-892, sheets 1-3

455-893

455-894

455-895, sheets 1-2

455-895, sheets 1-2

455-919

414-801, sheets 1-2

414-820

414-870

414-871, sheets 1-2

414-872, sheets 1-3

414-873

414-874

414-875

414-881, sheets 1-2

414-882, sheets 1-2

March 2017

Revision 18

List of Drawings (continued)

Revision No. Title

Rev4 Assembly, Transport Cask, NAC-MPC Rev 3 Spacers, Transpo11 Cask, MPC-Yankee

Rev 5 Canister Shell, MPC-Yankee

Rev 8 Details, Canister, MPC-Yankee

Rev 7P2 Details, Canister, MPC-Y ankee

Rev 12 Assembly, Transportable Storage Canister (TSC), MPC-Yankee

Rev 1 lPl Assembly, Transportable Storage Canister (TSC), MPC-Yankee

Rev4 Assembly, Drain Tube, Canister, MPC-Yankee

Rev 8 PWR Fuel Tube, MPC-Yankee

Rev4 Basket Assembly, 24 GTCC Container, MPC-Yankee

Rev 8 Assembly, Transportable Storage Canister (TSC), 24 GTCC Container, MPC-Yankee

Rev 1 Bottom Weldment, Fuel Basket, MPC-Yankee

Rev 2PO Bottom Weldment, Fuel Basket, MPC-Yankee

Rev 3 Top Weldment, Fuel Basket, MPC-Yankee

Rev 3PO Top Weldment, Fuel Basket, MPC-Yankee

Rev 3 Support Disk and Misc. Basket Details, MPC-Yankee

Rev 2 Heat Transfer Disk, Fuel Basket, MPC-Yankee

Rev 5 Fuel Basket Assembly, MPC-Yankee

Rev 5PO Fuel Basket Assembly, MPC-Yankee

Rev2 Retainer, United Nuclear Test Assy, MPC-Yankee

Rev 2 Cask Assembly, NAC-STC, CY-MPC

Rev 0 Canister Spacer CY-MPC

Rev 3 Canister Shell, CY-MPC

Rev 6 Details, Canister CY-MPC

Rev 6 Assembly, Transportable Storage Canister (TSC), CY-MPC

Rev 2 Drain Tube Assembly, CY-MPC

Rev 0 Shim, Canister, CY-MPC

Rev 0 Spacer Shim, Canister, CY-MPC

Rev4 Fuel Tube, Transportable Storage Canister (TSC), CY-MPC

Rev 4 Oversize Fuel Tube, Transportable Storage Canister (TSC), CY-MPC

1-vi

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NAC-STC SAR

Docket No. 71-9235

(4) Vent port coverplate interseal test hole threaded plug with metallic 0-ring;

(5) Drain port coverplate and the coverplate outer metallic 0-ring; and,

(6) Drain port coverplate interseal test hole threaded plug with metallic 0-ring.

March 2017

Revision 18

Metallic 0-rings are required for the storage configuration and are qualified for transport prior to

shipment in accordance with the operating procedure.

The NAC-STC is designed to meet IO CFR 71 and IAEA Safety Series No. SSR-6 licensing

requirements for spent fuel transport packages. The transport licensing requirements include

providing safe containment during the handling and transpo1t of spent nuclear fuel. Certain design

features of the NAC-STC that have been included for the sole purpose of satisfying storage

licensing requirements also provide added safety for transport conditions. The design features of

the NAC-STC include: inner and outer lids, redundant seals at each containment boundary

penetration, cavity penetrations located in the inner lid, and a puncture-resistant outer shell and

outer lid.

This Safety Analysis Report is written for transport cask licensing only. Design features related

to storage cask licensing are included for clarity and for ease of review.

The NAC-STC closure design provides dual lids for transport and storage operations, as well as

protection of the vent and drain ports that are located in the inner lid. This design permits

performance of a periodic verification leak test on the containment seals prior to transpo1t

following extended storage. Both the inner and outer lids are installed during transport and storage.

The inner lid and its 0-rings are the major removable components in the primary containment

boundary. Two concentric 0-rings are used to seal the inner lid to the cask cavity flange. An

0-ring test port connects to the annulus between the two 0-rings to permit leak testing.

The vent and drain port coverplates, which protect the vent and drain ports located in the inner lid,

are also part of the primary containment boundary of the cask. Each coverplate is sealed by two

concentric 0-rings.

As described in Section 4. I, the inner 0-rings of the inner lid and two coverplates are the

containment boundary for contents ( either directly loaded fuel or a loaded transportable storage

canister) that is loaded for transport without interim site storage. The outer 0-rings of these

I. 1-3

NAC-STC SAR

Docket No. 71-9235, Revision 17C

June 2018

Revision 18B

components are the containment boundary for directly loaded fuel that is to be transported after an

extended period of storage.

The inner lid and coverplate 0-rings may be either metallic or non-metallic as shown in the License

Drawings. However, metallic 0-rings must be used when the NAC-STC is directly loaded for

long-term storage or for the transport of canistered contents. The metallic 0-rings provide long

term sealing capability in an elevated temperature and radiation environment.

The outer lid provides a sealed secondary closure for transport and storage operations using a

single 0-ring. The 0-ring may be either metallic or non-metallic. The outer lid protects the inner

lid and the vent and drain ports from external puncture events.

There are two penetrations in the top forging: an interlid port, which serves primarily as a drain

for the inter! id region, and a pressure port, which may house a transducer that monitors the pressure

in the interlid region during storage. During transpo1t, the pressure port is closed by a threaded

plug. The pressure port plug is covered by the transport p01t coverplate. The interlid and pressure

port penetrations in the top forging are protected by SA-705, Type 630, 17-4 precipitation

hardened (PH) stainless steel po1t covers with two Viton 0-rings.

The body of the NAC-STC is a smooth right-circular cylinder of multiwall construction, consisting

of stainless steel inner and outer shells separated by lead gamma radiation shielding, which is

poured in place. The center section of the inner shell is fabricated from Type 304 stainless steel.

At each end of the inner shell center section, inner shell rings fabricated from Type XM-19

stainless steel provide the transition to the bottom inner forging and the top forging. The outer shell

is also fabricated from Type 304 stainless steel. The inner and outer shells are welded to the Type

304 stainless steel top forging, which is a ring that is machined to mate with the inner and outer

lids. The inner and outer shells are also welded to the Type 304 stainless steel bottom inner and

outer forgings, respectively. The cask bottom consists of the two forgings and a plate with neutron

shield material sandwiched between the bottom inner forging and the bottom plate. Neutron shield

material is also placed in an annulus that surrounds the cask outer shell along the length of the cask

cavity. The neutron shielding material is a solid synthetic polymer (NS-4-FR). The neutron shield

annulus is enclosed by a Type 304 stainless steel shell and by end plates that are welded to the

outer shell. Two pressure relief valves are provided in the bottom of the neutron shield annulus to

relieve pressure in the neutron shield annulus due to a severe thermal accident condition (fire).

Neutron shielding is also provided on the top of the cask by a layer ofNS-4-FR enclosed in the

inner lid.

1.1-4

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NAC-STC SAR

Docket No. 71-9235, Revision l 7C

June 2018

Revision 18B

The pressure port, located in the top forging, houses the transducer that monitors the pressure in the

interlid region during storage. The pressure transducer is removed during transpo1t. The interlid

po11 penetrates the top forging into the region between the inner and the outer lids and serves as a

drain for the interlid region and as a port to pressurize the interlid region for seal testing purposes.

The inter! id port is closed by a quick disconnect. The basic geometry of the inter! id and pressure

ports and port covers is identical. Each po11 has a 4.5-inch diameter opening that is a minimum of

1.1 inches deep. Concentric with the port opening is a 2.93-inch diameter bore. This bore acts as a

lead-in to the 2.875-inch diameter bore that serves as the sealing surface for the two Viton 0-rings

in the po11 cover.

Both of the port covers are fabricated from SA-705, Type 630, HJ 150, 17-4 PH stainless steel.

The port covers resemble a cup-shape and have the geometrical appearance of a thick round end

plate with a cylindrical body. The end plate of the port cover is 4.5 inches in diameter and 1.0

inch thick. The three 3/8 - 16 UNC port cover bolts, which are fabricated from SA-193, Grade

86, Type 410 stainless steel, are countersunk flush with the top of the port cover. There are two

Viton 0-rings on the cylindrical body of the port covers with a seal test port between the 0-rings.

A retainer is bolted to the open end of the cylindrical body of the port cover to retain the 0-rings

and the spacer between them after assembly. The port cover design permits the thick end plate to

absorb an impact, while any deflection of the end plate results in the 0-rings sliding in the bore of

the port with the seal maintained.

The basic geometry of the vent port and coverplate, and the drain port and coverplate, are identical

to each other. Each port has a 6.53-inch diameter opening in the inner lid that is 1.8 inches deep.

Concentric with the port opening is a 3.25-inch diameter bore that houses the 1.0-inch diameter

quick disconnect. As shown in Drawing 423-806, the 1.0-inch thick vent and drain port

coverplates are fabricated from SA-240, Type 304 stainless steel. When installed, the port

coverplates are recessed 0.8 inch below the top surface of the inner lid. The vent and drain poti

coverplates are sealed to the inner lid by the metallic or nonmetallic 0-rings on the bottom face of

each port coverplate. The four 1 /2 - 13 UNC port coverplate bolts are fabricated from SA-193,

Grade 86, Type 410 stainless steel. The bolt holes are countersunk so that the bolt heads are flush

with the top of the po11 coverplate. Metallic 0-rings are used for storage and for transport

following storage, and for transpo1i of canistered spent fuel and HL W without interim storage.

Either metallic or non-metallic 0-rings may be used for transport without interim storage after

loading. The outer metallic 0-ring provides the primary containment seal for transport after

storage, while the inner 0-ring (either metallic or non-metallic) provides the primary containment

seal for transport without interim storage after loading .

1.2-9

NAC-STC SAR

Docket No. 71-9235

1.2.1.2.5 Lifting Trunnions and Rotation Trunnion Recesses

March 2017

Revision 18

The NAC-STC has four lifting trunnions that are fabricated from SA-705, Type 630, Hl 150,

17-4 PH stainless steel and are welded into 2.0-inch deep recesses in the top forging at 90-degree

intervals around the cask circumference. Only two diametrically opposite lifting trunnions are

required to lift the NAC-STC. The lifting trunnions are 5.5 inches in diameter and have a

load-bearing width of 2.5 inches. The trunnions are machined to create a 0.38-inch thick end

flange, which acts as a safety stop to ensure proper engagement and to prevent inadvertent

disengagement of the lifting yoke.

There are two rotation trunnion recesses located near the bottom end of the NAC-STC. The

rotation trunnion recesses are located approximately 18 inches above the bottom of the cask in line

with two of the lifting trunnion, but 3.0 inches offset from the cask centerline to ensure that rotation

of the cask occurs in the proper direction. Each recess is fabricated from SA-705, Type 630, 17-4

PH stainless steel and is groove-welded to the bottom outer forging. The recess is 6.0 inches

square and 4.13 inches deep, with a full radius at the top of the recess that engages with the rotation

support. The neutron shield shell is cut out to accommodate the rotation trunnion recesses.

1.2.1.2.6 Transport Impact Limiters

The NAC-STC is equipped with removable, cup-shaped impact limiters that are bolted over each

end of the cask to ensure that the design impact loads for the cask are not exceeded for any of the

defined normal operation or accident drop conditions. The NAC-STC transport impact limiters

are provided in two configurations. The standard configuration is constructed of a combination of

redwood and balsa wood and is referred to as the redwood impact limiter. The other configuration

is constructed using only balsa wood and is referred to as the balsa impact limiter. Both

configurations are completely enclosed in a stainless steel shell. The upper impact limiter has

cutouts in its inside diameter for clearance with the lifting trunnions. The impact limiters absorb

the energy of a cask drop by crushing the redwood and/or balsa wood. The force required to crush

the impact limiter is determined by the amount and location of the wood and its grain direction.

The upper and lower impact limiters are bolted over each end of the cask body by 16 equally

spaced attachment rods and nuts. The lightweight impact limiters have a lower weight and

improved crush characteristics compared to the standard impact limiters, and accommodate a

higher cask content weight and higher cask total weight. As shown in Section

1.2-10

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NAC-STC SAR


Table 1.2-1 Design Characteristics of the NAC-STC (continued)

Design Characteristics Dimension 1

Seals (0-rings) for Spent Fuel Storage Configuration Prior to Transport, and Immediate Transport Configuration for Canister Spent Fuel, GTCC Waste, and HL W Overpacks

- Inner Lid

- Inner 0.25 dia. X n.251 dia. Metal Seal

- Outer 0.25 dia. x 73.497 dia. Metal Seal

- Port Coverplates

- Inner 0.125 dia. x 3.875 dia. Metal Seal

- Outer 0.125 dia. x 4.500 dia. Metal Seal

- Outer Lid 0.250 dia. x 82.060 dia. Metal Seal

- Port Covers

- Primary 0.103 dia. x 2.675 dia. Viton

- Secondary 0.103 dia. x 2.675 dia. Viton

1.2-43

June 2018

Revision 18B

Material

NAC-STC SAR


June2018

Revision 18B

Table 1.2-1 Design Characteristics of the NAC-STC (continued)

Design Characteristics Dimension 1 Material

Seals (0-rings) for Immediate Spent Fuel Transpo11 Configuration

- Inner Lid

- Inner 0.25 dia. x 72.251 dia. Viton or Metal Seal

- Outer 0.25 dia. x 73.497 dia. Viton

- Port Coverplates

- Inner 0.125 dia. x 3.875 dia. Viton or Metal Seal

- Outer 0.125 dia. x 4.500 dia. Viton

- Outer Lid 0.250 dia. x 82.060 dia. Viton

- Port Covers

- Primary 0.103 dia. x 2.675 dia. Viton

- Secondary 0.103 dia. x 2.675 dia. Viton I. Dimensions in Inches unless otherwise noted.

1.2-44

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•

Drawing Nos. Drawing 423-209, Revision 02; Drawing 423-210, Revision 02; and Drawing 423-870, Rev 8 have been withheld as Sensitive Unclassified Non-Safeguards Information pursuant to 10 CFR 2.390.

•

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NAC-STC SAR

Docket No. 71-9235, Revision I 7C

June 2018

Revision I SB

bottom connects the inner and outer shells, providing for the bottom end closure, as well as both

gamma and neutron radiation shielding in the axial direction.

The inner lid, bolts, and 0-rings are the primary closure components of the NAC-STC for

transport conditions. The outer lid and 0-ring provide a secondary closure boundary.

The vent po1t and the drain port are located in the inner lid and are each protected by a port

coverplate. The primary containment boundary at the vent port and at the drain po1t is the port

coverplate and its 0-rings. The 0-ring is located in the bottom surface of the port coverplate. A

second 0-ring is also located in the bottom surface of the port coverplate, inside of, and

concentric with, the first 0-ring.

The forty-two 1 I /2 - 8 UN inner lid bolts are preloaded by an installation torque to restrain

rotation of the edge of the inner lid and to maintain a containment seal for the critical load

condition. This condition is a uniformly distributed pressure resulting from the impact of the

basket and cavity contents on the inner surface during a top end or top corner impact. The

critical design load condition for the inner lid bolts, as listed in Table 2.7.1.6-2, Section 2.7.1.6,

is a 54.7 g top corner impact (IO CFR 71 Hypothetical Accident Condition). The critical design

load condition for the inner lid is the top end impact, Section 2.7.1.6.

The outer lid is bolted to the top forging by the thirty-six 1 - 8 UNC outer lid bolts, which are

installed to a specified torque. The torque provides a total bolt preload that exceeds the

maximum applied bolt load for the critical load condition, preventing any lid and 0-ring

movement that might result in a loss of secondary seal integrity. The critical design load

condition for the outer lid bolts, as listed in Table 2.7.1.6-4, Section 2.7.1.6, is a 51.3g side

impact (10 CFR 71 Hypothetical Accident Condition). The critical design load condition for the

outer lid is the pin puncture accident condition. The NAC-STC outer lid bolts are loaded by the

interlid region pressure, the 0-ring compression force, and by either the impact limiter crush

force during a top end or top corner impact, or by a concentrated center load during a pin

puncture impact. The outer lid seal is provided by an 0-ring, which is tested by pressurizing the

interlid region.

In addition to the main closure, the secondary closure boundary of the NAC-STC also includes

the two po1ts located in the top forging-the interlid port and the pressure port. Each of these

ports is protected and sealed by a recessed, bolted port cover with two Viton 0-rings. The port

covers are installed with new 0-rings just prior to transport (a slightly different port cover is

2.1.1-3

NAC-STC SAR


June 2018

Revision l 8B

installed during storage operation). The seal at each port cover is verified by pressure-testing the

annulus between the two Viton 0-rings.

The neutron shielding material, NS-4-FR, is a solid synthetic polymer that absorbs the neutron

radiation emitted by the cask contents. In addition to the radial neutron shielding along the cask

length, neutron shielding is provided in the axial direction at each end of the cask by circular

layers ofNS-4-FR enclosed in the inner lid and in the cask bottom.

Four external trunnions are welded to the top forging of the NAC-STC at 90-degree intervals

around the circumference of the cask. These trunnions are provided for lifting and handling the

cask. Either a redundant (four trunnions) or a nonredundant (two trunnions) lifting system may

be used. However, each pair of opposing trunnions are conservatively designed to satisfy the

heavy lifting requirements of NUREG-0612 for a nonredundant lift, as well as the requirements

of 10 CFR 7 l .45(a) and paragraph 607 of IAEA Safety Standards Series No. SSR-6. Two

rotation trunnion recesses are welded to the bottom outer forging near the bottom of the cask.

The neutron shield is cut out to accommodate the placement of the rotation trunnion recesses,

which are used to attach the bottom of the cask to the transport vehicle and to rotate the cask

from the vertical lifting position to the horizontal position and vice-versa.

As discussed above, two transport impact limiter configurations are used with the NAC-STC

cask to limit the impact loads that may act on the cask. The impact limiters absorb the energy of

a cask drop impact through the crushing of the wood in the limiters. A balsa impact limiter

design must be used when the NAC-STC is transporting spent fuel or GTCC waste in the

CY-MPC canister configuration. When transpo1ting directly loaded fuel or the Yankee-MPC

canister, either the redwood or balsa impact limiter configuration may be used.

The NAC-STC fuel basket is constructed of stainless steel and has a capacity of 26 PWR fuel

assemblies. The fuel basket has a cylindrical shape with a series of support disks that provide

lateral support for the square, stainless steel fuel tubes, which encase neutron absorber sheets or

plates on each of the four sides. The support disks are separated and supported at 4.87-inch

intervals by a threaded rod and spacer nuts at six locations. Aluminum heat transfer disks are

located in the central region of the fuel basket and are supported by the six threaded rods and

spacer nuts. The stainless steel support disks have adequate strength at the basket temperatures

that occur during the transport and/or storage of 26 design-basis PWR fuel assemblies.

For the Yankee Class fuel and GTCC waste, the Yankee-MPC transportable storage canister

(canister) serves as the enclosure of the spent fuel assemblies, damaged fuel cans and GTCC

2.1.1-4

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NAC-STC SAR

Docket No. 71-9235

2.3.2 Austenitic Stainless Steels

March 2017

Revision 18

The primary structural components of the NAC-STC body, excluding: (1) the inner shell rings

(transition sections of the inner shell); (2) the outer lid; (3) the lifting trunnions; and (4) the

rotation trunnion recesses, are fabricated from Type 304 stainless steel. In addition to the cask

body components fabricated from Type 304 stainless steel, the fuel tubes and fuel basket top and

bottom weldment plates are fabricated from the same material. This material is selected because

it is strong, ductile, and highly resistant to corrosion and brittle fracture. Type XM-19 stainless

steel is selected for the inner shell rings at the ends of the inner shell because the high strength of

Type XM-19 stainless steel provides additional resistance to shear buckling in those sections of

the inner shell.

The mechanical properties of SA-240 (plate), Type 304 stainless steel are tabulated in Table

2.3.2-1. The mechanical properties of SA-336 (forging), Type 304 stainless steel are tabulated in

Table 2.3.2-2. The mechanical properties of SA-240 (plate), Type XM-19 stainless steel are

tabulated in Table 2.3.2-3 .

The primary structural components of the Yankee-MPC and CY-MPC canisters and baskets,

excluding the support disks and heat transfer disks, are fabricated from Type 304 and Type 304L

stainless steels. This material is selected because it is strong, ductile, and highly resistant to

corrosion and brittle fractures. The associated mechanical properties of the Type 304 and 304L

stainless steels are tabulated in Tables 2.3.2-1, 2.3.2-2 and 2.3.2-4 .

2.3.2-1

NAC-STC SAR

Docket No. 71-9235

Table 2.3 .2-1 Mechanical Properties of SA 240, Type 304 Stainless Steel

Temperature (°F)

Property (units) -40 -20 +70 +200 +300

Ultimate Strength 1 75.0 75.0 75.0 71.0 66.0 (ksi)

Yield Strength2 30.0 30.0 30.0 25.0 22.5 (ksi)

Design Stress 20.0 20.0 20.0 20.0 20.0 Jntensity3 (ksi)

Modulus of 28.7E+3 28.7E+3 28.3E+3 27.6E+3 27.0E+3 Elasticity4 (ksi)

Alternating Stress5 718.0 718.0 708.0 690.5 675.5 (iv, IO cycles (ksi)

Alternating Stress5 28.7 28.7 28.3 27.6 27.0 (a), 106 cycles (ksi)

Coefficient of 8.13E-6 8.19E-6 8.46E-6 8.79E-6 9.00E-6 Thermal Expansion6

(in/inl°F)

Poisson's Ratio7 0.31

Density (lbm/in3 ) 497 lbm/ft3 (0.288 lbm/in3)

1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-I. 5 "ASME Boiler and Pressure Vessel Code," Section Ill, Appendix I, Table 1-9.1. 6 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-I. 7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-I.

2.3.2-2

+400

64.4

20.7

18.7

26.5E+3

663.0

26.5

9.19E-6

June 2018

Revision 18B

+500 +750

63.5 63.1

19.4 17.3

17.5 15.6

25.8E+3 24.4E+3

645.5 610.4

25.8 24.4

9.37E-6 9.76E-6

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NAC-STC SAR

Docket No. 71-9235, Revision 17B


Temperature (°F)

Property (units) -40 -20 +70 +200 +300

Ultimate Strength 1 70.0 70.0 70.0 66.2 61.5 (ksi)

Yield Strength2 30.0 30.0 30.0 25.0 22.5 (ksi)

Design Stress 20.0 20.0 20.0 20.0 20.0 Intensity3 (ksi)

Modulus of 28.7E+3 28.7E+3 28.3E+3 27.6E+3 27.0E+3 Elasticity4 (ksi)

Alternating Stress5 718.0 718.0 708.0 690.5 675.5 @ IO cycles (ksi)

Alternating Stress5 28.7 28.7 28.3 27.6 27.0 @ I 06 cycles (ksi)

Coefficient of 8.13E-6 8.19E-6 8.46E-6 8.79E-6 9.00E-6 Thermal Expansion6

(in/in/°F)


Density8 (lbm/in3) 497 lbm/ft3 (0.288 lbm/in3)

1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section II, Pmi D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-I.

"ASME Boiler and Pressure Vessel Code," Section III, Appendix I, Table 1-9.1. 6 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-I. 7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-!. 8 "Nuclear Materials Handbook," Volume I, Design Data, Property Code 3304 .

2.3.2-3

+400

60.0

20.7

18.7

26.5E+3

663.0

26.5

9.19E-6

June2018

Revision 18B

+500 +750

59.3 58.9

19.4 17.3

17.5 15.6

25.8E+3 24.4E+3

645.5 610.4

25.8 24.4

9.37E-6 9.76E-6

NAC-STC SAR

Docket No. 71-9235

Table 2.3.2-3 Mechanical Properties of Type XM-19 Stainless Steel

Temperature (°F)

Property (units)9 -40 -20 +70 +200 +300 +400

Ultimate Strength 1 100.0 100.0 100.0 99.5 94.3 90.7 (ksi)

Yield Strength2 55.0 55.0 55.0 47.0 43.4 40.8 (ksi)

Design Stress 33.3 33.3 33.3 33.2 31.4 30.2 lntensity3 (ksi)

Modulus of 28.3E+3 28.3E+3 28.3E+3 27.0E+3 27.0E+3 26.5E+3 Elasticity4 (ksi)

Alternating Stress5 708.0 708.0 708.0 690.5 675.5 663.0 (]iJ, 10 cycles (ksi)

Alternating Stress5 28.3 28.3 28.3 27.6 27.0 26.5 (]iJ, 106 cycles (ksi)

Coefficient of 8.13E-6 8.l 9E-6 8.46E-6 8.79E-6 9.00E-6 9.19E-6 Thermal Expansion6

(in/in/°F)


Density8 (lbm/in3) 497 lbm/ft3 (0.288 lbm/in3)

"ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1. 3 "ASl\!lE Boiler and Pressure Vessel Code," Section II, Part D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-I. 5 "ASME Boiler and Pressure Vessel Code," Section Ill, Appendix I, Table 1-9.1. 6 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-I. 7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-1. 8 "Nuclear Materials Handbook," Volume I, Design Data, Property Code 3304.

June 2018

Revision l 8B

+500 +750

89.1 85.7

38.8 35.8

29.7 28.5

25.8E+3 24.4E+3

645.5 610.4

25.8 24.4

9.37E-6 9.76E-6

9 SA-182, FXM-19 stainless steel may be substituted for SA-240 XM-19 stainless steel provided that the SA-182 material yield and ultimate strengths are equal to or greater than those of the SA-240 material. The SA-182 forging material and the SA-240 plate material are both XM-19 austenitic stainless steels. Austenitic stainless steels do not experience a ductile-to-brittle transition for the range of temperatures considered in this Safety Analysis Report. Therefore, fracture toughness is not a concern.

2.3.2-4

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NAC-STC SAR


2.4.4.2.3 Shielding Materials

June 2018

Revision 18B

The primary shielding materials used in the NAC-STC-lead and NS-4-FR-are completely

enclosed and sealed in stainless steel. As previously described, there are no potential reactions of

these materials with the stainless steel or with the copper fins.

Therefore, there are no potential reactions associated with the cask shielding materials.

2.4.4.2.4 Criticality Control Material

The criticality control material is a sheet consisting of boron carbide mixed m an aluminum

alloy. This material is effectively a sheet of aluminum that is in contact with the aluminum alloy

fuel tubes and is exposed to the cask cavity environment. This material is protected by an oxide

layer that formed shortly after fabrication. The existing oxide layer effectively precludes fu11her

oxidation of the aluminum. Consequently, there are no potential reactions associated with the

aluminum-based criticality control material.

2.4.4.2.5 Energy Absorbing Material

The NAC-STC utilizes redwood and balsa wood for energy absorption in the impact limiters.

The wood is completely enclosed (sealed) in stainless steel and there are no potential reactions

between the wood and the stainless steel shells. The wood may be coated with a preservative

prior to installation in the impact limiter shell and blocks of wood may be glued together with an

epoxy adhesive. These are standard applications of preservatives and adhesives, so no post

application reactions will occur.

There are no potential reactions associated with the energy absorbing material.

2.4.4.2.6 Cellular Foam and Insulation

The NAC-STC utilizes layers of expansion foam and strips of insulation in the solid neutron

shield regions. The expansion foam permits thermal expansion of the solid neutron shield

material during normal operation, and the insulation protects the expansion foam during final

closure welding of the neutron shield shell to the end plate. The foam and the insulation are

nonflammable, nontoxic and noncorrosive silicone products that are used in the casks in a

standard design application .

2.4.4-7

NAC-STC SAR


June2018

Revision 18B

There are no potential reactions associated with the silicone expansion foam or insulation.

2.4.4.2.7 Lubricant and Grease

The dry film lubricants used with the NAC-STC meet the performance and general

compositional requirements of the nuclear power industry. One example is NEVER-SEEZ®

lubricant, which can be used on rotating bearing surfaces. Another example is Neolube®, which

can be used on threaded/mechanical connection surfaces. In addition, Dow Corning High

Vacuum Grease is an example of what can be used as an adherent/lubricant to lubricate and

retain the 0-ring seals in their grooves. None of these example lubricants contain elements or

compounds prohibited by the NRC. NEVER-SEEZ® is a superior, high temperature, anti-seize

and extreme pressure lubricant that contains flake particles of pure nickel, graphite and other

additives in a special grease carrier. Neolube® is 99% pure furnace graphite particles in

isopropanol. It has excellent radiation resistance and high chemical purity. It dries as a thin,

non-corrosive film with excellent adhesion, does not migrate, and is non-freezable. Dow

Corning High Vacuum Grease is a stiff, nonmelting, nonoxidizing, non gumming silicone

lubricating material that is insoluble in most solutions. There are no potential reactions

associated with these lubricants or grease. Other lubricants and greases maybe used as

alternatives to the examples given provided they meet the performance and general

compositional requirements of the nuclear power industry.

2.4.4.2.8

The NAC-STC utilizes seals formed from silicone rubber and Viton. Viton is a silicon

elastomer. Elastomer 0-rings are used for transport cask applications because of their excellent

short-term sealing capabilities, ease of handling, and more economical cost. All of the seal and

gasket materials have stable, non-reactive compositions. There are no potential reactions

associated with the NAC-STC seal materials.

2.4.4.3 General Effects of Identified Reactions

No significant potential galvanic or other reactions have been identified for the NAC-STC. The

only potential chemical reaction identified for the NAC-STC is that of aluminum with the spent

fuel pool water. As discussed in Section 2.4.4.2.2, it is possible at higher temperatures (above

I 50-l 60°F) that a flammable concentration of hydrogen might be generated by the

2.4.4-8

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NAC-STC SAR

Docket No. 71-9235

March 2017

Revision 18

aluminum/water reaction and accumulate beneath the canister shield lid during the canister

closure operations. The danger of potential ignition of the hydrogen is precluded by the

operating controls and procedures presented in Chapter 7. Therefore, no adverse conditions can

result during any phase of cask operations for normal, off-normal, or accident conditions.

2.4.4.4 Adequacy of the Cask Operating Procedures

Based on the results of this evaluation which resulted in only one identified reaction, aluminum

in pool water, it is concluded that the NAC-STC operating controls and procedures presented in

Chapter 7 are adequate to minimize the occurrence of hazardous conditions.

2.4.4.5 Effects of Reaction Products

No significant potential chemical, galvanic, or other reactions have been identified for the NAC

STC. Therefore, the overall integrity of the cask and the structural integrity and retrievability of

the spent fuel is not adversely affected for any cask operations throughout the design basis life of

the cask. Based on the evaluation, there will be no change in the cask or fuel cladding thermal

properties, and there will be no binding of mechanical surfaces, no change in basket clearances,

and no degradation of any safety components either directly or indirectly, since there are no

significant reactions identified .

2.4.4-9

NAC-STC SAR


June 2018

Revision l 8B

Table 2.4-1 Summary of NAC-STC Materials Categories and Operating Environments

ITEM MATERIAL ENVIRONMENT

Stainless Steels/ Alloys 304, 304L, XM-19, 17-4PH, Sealed Internal Ni Alloy, 410 Open Internal/ External

Nonferrous Metals ASTM B 152 Cu, Sealed Internal

606 l -T65 l Aluminum Alloy Open Internal/External

Shielding Materials NS-4-FR, Chemical Copper Enclosed Grade Lead

Criticality Control Materials Boroncarbide Enclosed

Aluminum 1100

Energy Absorbing Materials Balsa Wood, Redwood Enclosed

Cellular Foam/Insulation Silicone (HT-810 & 800), Enclosed Silicone Caulk (Dow Corning)

Lubricants & Greases Never-Seeze® Sealed Internal

Neolube® Open Internal

High Vacuum Grease® by Dow Corning

Seals & Gaskets Silicone Rubber, Viton Sealed Internal

Open Internal/ External

2.4.4-10

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NAC-STC SAR


June 2018

Revision 18B

Bolt Stress Evaluation

The maximum preload on the inner lid bolts in combination with the internal pressure force on

the lid, the 0-ring compression force on the lid, thermally induced loads and the inertial load of

the inner lid and cask contents due to the normal conditions of transport I-foot drop must not

exceed the allowable strength of the inner lid bolts.

A complete range of impact orientations is evaluated, from an end impact at 0° to a side impact at

90°, using 5° increments. A design acceleration of 20g is considered for all impact orientations

in the bolt evaluation for normal conditions for both impact limiter designs (redwood and balsa).

The cask contents weight for the CY-MPC is used since it bounds the cask content weight of the

Yankee-MPC configuration. The details of this evaluation are described and an example

calculation is provided in Section 2.10.8.2 for the hypothetical accident condition. Normal

conditions of transport results are summarized in Tables 2.6.7.5-1 and 2.6.7.5-2, corresponding

to a "hot" condition and a "cold" condition, respectively. The hot condition bolt temperatures are

assumed to be 200°F, as summarized in Table 3.4-5. The cold condition bolt temperature is

assumed to be -20°F, in accordance with regulatory requirements. Physical properties for the

SB-637 Grade N07718 nickel alloy bolts are conservatively taken at 270°F for both hot and cold

conditions. As defined in Table 2.1.2-1, the allowable maximum bolt stress for normal

conditions for primary membrane stress is two times the design stress intensity, 2Sm, resulting in

an allowable direct tension stress of 94.5 ksi at 270°F. As shown in Tables 2.6.7.5-1 and 2.6.7.5-

2, the total bolt stress is calculated to be less than the allowable stress for normal conditions of

transport. The minimum margin of safety is +0.06.

Bolt Thread Engagement Evaluation

The ultimate load capacity of the inner lid bolt/top forging threaded connection relative to the

ultimate tensile load capacity of the inner lid bolt is evaluated to ensure that the length of

engagement is sufficient to develop the full strength of the bolt. The inner lid bolt holes have

threaded inserts to protect the threads during the installation and removal of the bolts.

Component Description

Inner lid Bolt 11/2-8UN

SB-637, Grade N07718 Nickel Alloy Steel Bolting Material

Length in cask body= 9.75 inches

Su= 174.7 ksi at 270°F

2.6.7.5-5

NAC-STC SAR


Threaded Insert Helicoil #4190-24 CN x 2.50

(AMS 7245) 18-8 Stainless Steel

Length of insert= 2.50 inches

O.D. = I 5/8 - 8 UN Thread

Su= 200.0 ksi

Top Forging (Cask Body) Type 304 Stainless Steel

Thread depth= 3.0 - 0.125 = 2.875 in

Su = 62.9 ksi at 270°F

Bolt Strength

Tensile Area, At

Tensile Strength, Su

Bolt-Tensile Load Capacity, PBLC

= 1.492 in2 (I 1/2 - 8 UN Thread)

= 174.7 ksi

(1.492)(174,700)

260,650 lbs

Threaded Insert/Bolt Interface

Thread Size = 1 l/2-8UN

Engaged Length = Lbolt - hvasher - tiid - dset-down = 2.20 in

where:

Lbolt Length of bolt= 9.75 in

twasher Thickness of washer= 0.315 in [8 mm]

tiid = Thickness of lid at bolt location= 7.10 in dset-down Insert set-down distance= 0.14 in

External (Bolt) Thread Shear Area

where:

ASs = (n)(n)(Le)(Knmax)[(l /211) + (0.57735)(Esmin - Knmax)f

= 5.660 in2

n = 8 threads/in

Le = 2.20 in

Esmin = I .4093 in Min. Pitch Diameter of External Threads

Knmax = 1.390 in Max. Minor Diameter ofJnternal Threads

* FED-STD-H28 (1963), Page 103.

2.6.7.5-6

June 2018

Revision 18B •

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NAC-STC SAR

Docket No. 71-9235, Revision l 7B

Bolt Thread-Tensile Load Capacity, PBT

Internal (Insert) Thread Shear Area

= (5.660)(0.5* X 174,700)

= 494,400 lbs

June2018

Revision l 8B

ASn = (n)(n)(Le)(Dsmin)[(l/2n) + (0.57735)(Dsmin - Enmax)]**

= 7.704 in2

where:

Dsmin = 1.4828 in Min. Major Diameter of External Thread

Enmax = 1.4283 in Max. Pitch Diameter of Internal Thread

Insert Thread-Tensile Load Capacity, Pm

Threaded Insert/Top Forging Interface

Thread Size = I 5/8 - 8 UN

Engaged Length

Insert: Su

Top Forging: Su

= 2.20 in

= 200 ksi

= 62.9 ksi

External (Insert) Thread Shear Area

= (7.704) (0.5* X 200,000)

= 770,400 lbs

ASs = (rc)(n)(Le)(Knmax)[(]/2n) + (0.57735)(Esmin - Knmax)]**

= 6.164 in2

* Shear Strength Conservatively Assumed= (0.5)(Tensile Strength).

** FED-STD-H28 ( 1963), page 103 .

2.6.7.5-7

NAC-STC SAR


where:

n = 8 threads/inch

Le = 2.20 in

Esmin = 1.5342 in Min. Pitch Diameter of External Threads

Knmax = 1.515 in Max. Minor Diameter of Internal Threads

Insert Thread-Tensile Load Capacity, Pno = (6.164)(0.5* x 200,000)

= 616,400 lbs

Internal (Top Forging) Thread Shear Area

June 2018

Revision 18B

ASn = (n)(n)(Le)(Dsmin)[(l/2n) + (0.57735)(Dsmin - Enmax)]**

= 8.343 in2

where:

Dsmin = 1.6078 in Min. Major Diameter of External Thread

Enmax = 1.5535 in Max. Pitch Diameter oflnternal Thread

Top Forging Thread - Tensile Load Capacity, PTFT = (8.343)(0.5* x 62,900)

= 262,390 lbs

Component

Inner Lid Bolt

Bolt Thread

Insert I.D. Thread

Insert O.D. Thread

Top Forging Thread

Ultimate Load Capacity

(lbs)

260,6,50

494,400

770,400

616,400

262,390

Since the mm1mum Tensile Load Capacity of the threaded joint (262,390 lbs) exceeds the

maximum Tensile Load Capacity of the inner lid bolt (260,650 lbs), the load capacity of the

inner lid bolt is the controlling load capacity of the joint strength, and the design requirements

are satisfied. The inner lid bolt threaded-joint design is satisfactory.

* Shear Strength Conservatively Assumed = (0.5)(Tensile Strength).

** FED-STD-H28 (1963), page 103.

2.6.7.5-8

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NAC-STC SAR


2.6.11 Fabrication Conditions

June 2018

Revision 18B

The process of manufacturing the NAC-STC can introduce thermal stresses in the inner and outer

shells as a result of pouring molten lead between them. These thermal stresses are evaluated in

this section to provide assurance that the manufacturing process does not adversely affect the normal

operation of the cask or its ability to survive an accident. Any residual stresses in the containment

vessel shell due to inelastic strain associated with the secondary local bending stresses, which result

from the lead pour thermal gradient, must be considered in the total stress range for normal and

accident load conditions according to Regulatory Position 7 of Regulatory Guide 7.6. Residual

stresses in the containment vessel and the outer shell induced by shrinkage of the lead shielding

after the lead pouring operation are relieved early in the life of the cask because of the low creep

strength of lead.

For the lead pouring process, the temperatures of the cask shells are controlled between 640°F

(338°C) and 740°F (393°C), and the maximum lead temperature before pouring is 790°F (421 °C).

Heating of the cask is performed using heaters inside the inner shell and heating rings around the

outside of the outer shell. Heat up is time controlled, consistent with maintaining shell

temperatures uniformly. The heating procedures ensure that the surface temperature of the cask

does not exceed 800°F (427°C). The shell temperatures are measured by thermocouples attached

to the shell surfaces. A portable thermometer is also used to measure temperature at any location.

Cask heating is carried out after all of the preparations have been completed (including melting of

the lead) in order to minimize the time that the cask is at elevated temperatures.

The lead is poured after the cask reaches the specified temperatures. Prior to lead pouring, the cask

flange area is heated with hand-held burners to between 640°F (338°C) and 740°F (393°C).

Pouring is carried out continuously using a filling tube with its open end maintained under the lead

surface. The pouring time is kept as short as possible. During pouring, the interior heaters and

exterior heating rings are continuously energized.

The cooling process consists of sequentially turning the exterior heating rings and interior heaters

off, starting from the lowest point, and of spraying the cask with water from the outside. A layer

of molten lead is maintained until the upper surface starts to solidify. This process allows the molten

lead to fill the open space below it created by the lead shrinkage as it cools.

The basic requirements and procedures for the NAC-STC lead pour operations are described in

Section 8.4.2 and 8.4.3 .

2.6.11-1

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NAC-STC SAR


2.6.11.1 Lead Pour

2.6.11.1.1 Cask Shell Geometry

At 70°F, the Type 304 stainless steel shell geometry is as follows:

Inner Shell

Inside Diameter (di)

Outside Diameter (do)

Shell Thickness (ti)

Outer Shell

Inside Diameter (Di)

Outside Diameter (Do)

Shell Thickness (To)

= 71.0 in

= 74.0 in

= 1.5 in

= 81.4 in

= 86.7 in

= 2.65 in

2.6.11.1.2 Stresses Due to Lead Pour

June 2018

Revision l 8B

As stated in the Lead Pour Procedures in Sections 8.4.2 and 8.4.3, the maximum lead temperature

during the pouring operations is 790°F. Assuming that the lead and the inner and outer shells are

uniformly at 790°F, the hydrostatic pressure produced by the column of lead is:

p = ph

= 66 psi

where

p = 0.41 lb/in3 (lead density)

h = 161 in (maximum height of lead column)

2.6.11.1-1

NAC-STC SAR


At 790°F, the shell geometric dimensions are:

where

d~ = do ( 1 + a ~ T)

n; =Di(l+a~T)

t' = t (1 + a ~ T)

a = J 0.09 x 1 o-6 in/in/°F at 790°F (stainless steel)

~T = 790 - 70 = 720°F

d~ = (74.0) [1 + (1 o.o9 x I o-6)(720)]

= 74.54 in

d~ = [71.o;

74·0

] [I+ (10.09 x 10-6)(720)]

= 73.03 in

n; = (81.4)[1 + (10.09 X 10-6)(720)]

= 81.99 in

, [81.4 + 86.7] Dm= 2

[(1+(10.09xl0-6)(720)]

= 84.66 in

2.6.11.1-2

June 2018

Revision 18B

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NAC-STC SAR


' ti = (1.50) [I+ (10.99 x 10-6)(720)]

= 1.51 in

' t 0 = (2.65) [1 + (I 0.99 X 1 o-6)(720)]

= 2.669 in

June 2018

Revision 18B

The inner shell is subjected to an external hydrostatic pressure, and the outer shell to an internal

hydrostatic pressure, of 66 psi. This causes the inner shell to decrease in diameter and the outer

shell to increase in diameter.

The inner shell decreases in size radially (Roark and Young, 5th ed., Case 1 b, page 448):

where

b.rm

(-66)(74.54/2)2 - 0.00251 in (24.17 x10 6 )(1.51)

(-66)(73.03/2)2 - 0.00241 in (24.17 xl0 6 )(1.51)

E = 24.17 x 106 psi at 790°F.

The outer shell increases in size radially:

q(D{,,/2) 2

Et0

(66

)(8

1.9912)

2 = 0.00172 in (24.17 x106 )(2.669)

(66)(84.66/2)2 = 0.00183 in (24.17 x10G )(2.669)

2.6.11.1-3

NAC-STC SAR


The shell geometries at 790°F and 66 psi hydrostatic pressure are:

d"o = 74.54 - (2)(0.00251) = 74.535 in

D"i = 81.99 + (2)(0.00172) = 81.993 in

d\n = 73 .03 - (2)(0.00241) = 73 .025 in

D"m = 84.66 + (2)(0.00183) = 84.666 in

June 2018

Revision 18B

The hoop stresses are evaluated at the mean diameter of the inner and outer shells at 790°F:

= Pd;~ 2t'

I

C-66)C73·025

) = -1596 psi (inner shell) (2)(1.51)

= PD;·n

2t~

C66)CB 4·55

) = 104 7 psi ( outer shell) (2)(2.669)

These stresses are negligible.

2.6.11.1-4

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NAC-STC SAR

Docket No. 71-9235, Revision I 7B

2.6.11.2 Cooldown

2.6.11.2.1 Hoop Stresses

June 2018

Revision I 8B

Lead decreases in volume during solidification. As the lower lead region solidifies, the molten lead

above fills the shrinkage void between the solidifying lead and the inner and outer shells, thus,

maintaining the 66 psi pressure on the shells.

The stress-free inner and outer radii of the solidified lead can be calculated (Roark and Young, 5th

ed., Cases I a and 1 c, page 504) as:

L'ia - q [ 2ab2 l qa [a 2 - b2 - v] = Eq a(v - I) E a 2 -b2 E a 2 -b2

=-0.001104 in

L'ib + v - - = - b(v - I) = q b [ a 2

+ b 2 l q [ 2a

2 b l q

E a 2 -b2 E a 2 -b2 E

= -0.001004 in

where

q = 66 psi (pressure)

E = 1.47 x 106 psi at 620°F (modulus of elasticity)

v = 0.4 (Poisson's ratio)

a = D"i/2 = 81.993/2 = 40.9965

b = d"o/2 = 74.535/2 = 37.2675

then

Roi = 40.9965 - 0.001104 = 40.9954

2.6.1 1.2-1

NAC-STC SAR


Ri1 = 37.2675 - 0.001004 = 37.2665

When cooled to 70°F, the inside radius of the lead is such that:

where

R;e = inside radius of the stress-free lead at 70°F

a = 20.4 x 1 o-6 in/in/°F

!-.T = 550°F (620-70)

then

likewise

_R_oe __ R' 1 + a!-.T - oe

R~e = 40.5405 in

June 2018

Revision 18B

The outside radius of the stress-free inner shell is 74.0/2 = 37.0 inches, which is larger than the

stress-free inner radius of the lead shell. Therefore, there exists an interface pressure between the

lead and the inner shell after cooling to 70DF. The interface pressure, when acting on the lead

cylinder and inner shell, is such that the inner radius of the lead cylinder is the same as the outer

radius of the inner shell (Roark and Young, 5th ed., Case 1 a, page 504).

2.6.11.2-2

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NAC-STC SAR


where

then

= b + q b [ a 2 + b 2 + v] E a 2

- b2

R;e = inside radius of lead cylinder at 70°F

V = 0.4

E = 2.28 x I 06 psi at 70°F

a = 40.5405 in

b = 36.853 in

= 36.853 + ( 36.853q ) ((40.5405)2

+ (36.853)2 + 0.4)

2.28 x10 6 (40.5405) 2 - (36.853) 2

= 36.853 + 1.765 x I o-4q

June 2018

Revision 18B

The outside radius of the inner shell at 70°F under the interface pressure, q, (Roark and Young,

5th ed., Case le, page 504) is:

where

ro = as -L'ias

= as _ qa s (a; + b; _ vJ E a 2

- b2 s s

ro = outside radius of inner shell at 70°F

as = 74.0/2 = 37.0 in

bs = 71.0/2 = 35.5 in

2.6.11.2-3

NAC-STC SAR


then

E = 28.3 x 106 psi at 70°F

V = 0.275

= _ ( 37.0q :( (37.0)2

+ (35.5)2

_ O 275] ro 37.0 6 2 2 · 28.3 X 10 (37.0) - (35.5)

= 37.0 - 3.)25 X 10-5q

Equating R;e and ro and solving for q:

q = 708 psi interface pressure

The lead shell geometry is:

((36.853)(708)) ((40.5405)

2 + (36.853)2

- 0.4) 2.28 Xl06 (40.5405)2 - (36.853) 2

= 36.969 in

= 40.5405 + ( (708) ) ( (2)( 40.5405)(36.853)2

) 2.28 Xl0 6 (40.5405) 2 - (36.853) 2

= 40.66 in

June 2018

Revision l 8B

The interference between the lead shell and the inner shell is 0.142 inch (37.0 - 36.853). To fully

accommodate this interference, the lead must undergo a strain of 0.172/36.853 = 0.004 or 0.4

percent. From Figure 24 ofNUREG/CR-0481, the lead stress for the above strain is 800 psi. The

corresponding interface pressure for this stress in the lead shell is:

2.6.11.2-4

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NAC-STC SAR


q [

, 2 , 2 J = (S) Roe - Ric , 2 , 2

Roe + Ric

= (800) ((40.660)2

- (39.969)2

) ( 40.660)2 + (39.969)2

= 76 psi interface pressure

The change in geometry of the inner shell for this interface pressure is:

L1a [ -76 ] [(2)(37.0)(35.5)2

] - 28.3 X 106 (37.0)2 - (35.5) 2

= 0.0023 in

This can conservatively be neglected in the analysis. The inner shell hoop stress is:

Shis = (-76) [(37.0)2 + (35.5)21 (37.0) 2 - (35.5)2

=-1837psi

This stress is negligible.

2.6.11.2.2 Axial Stresses

June 2018

Revision l 8B

Axial stresses also develop in the lead shell and inner shell during fabrication as a result of the

unequal shrinkage of the lead and steel shells. Assume bonding of the lead shell to the inner shell

during the cool down process after completion of lead pouring. The strain in the lead, when cooled

to 70°F, is:

Ee = (a.1 - Us)L1T

= 0.0060 in/in or 0.60 percent

where

2.6.11.2-5

NAC-STC SAR


a.e = 20.4 X 1 o-6 in/in/°F

a5 = 9.56 x 10-6 in/in/°F

LiT = 620 - 70 = 550°F

June 2018

Revision 18B

Extrapolating from Figure 24 ofNUREG/CR-0481 for this strain, an axial stress of approximately

825 psi exists in the lead shell. The total force in the lead caused by assuming no deformation of

the inner shell is:

PsPb = Pe Ae

= 825n[(40.7)2 - (37.0)2]

= 745,120 lb tensile force

The corresponding compression stress in the inner shell to maintain equilibrium is:

-745,120

n[(37.0) 2 - (35.5) 2

]

= -2180 psi

This stress is negligible.

This is a highly conservative estimate of the compressive stress that can develop in the inner shell

for the following reasons:

1. It assumes no axial deformation of the inner shell and no load development in the

outer shell.

2.6.11.2-6

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• NAC-STC SAR


June2018

Revision 18B

2. Creep in the lead is neglected. This also reduces the stress and force in the_ lead

(Section 2.6.11.3).

3. It assumes the strain is uniform through the thickness of the lead shell, i.e., no shear

strain exists in the plane formed by the radial axis and the longitudinal axis. A

particle away from the inner shell should develop Jess strain, consequently lower

stress, than a particle adjacent to the inner shell; this also reduces the total force in

the lead shell.

2.6.11.2.3 Effects of Temperature Differential During Cooldown

The preceding analyses assume that the inner and outer shells and the lead are always at the same

temperature at any time during the cooldown process. This assumption may not be true under

actual conditions. However, because of the high thermal conductivity of the stainless steel and the

lead and because of the time-controlled cooldown process, the temperature differential between

any two of the above shells is kept to a minimum.

• If the inner shell is cooler than the lead, the interference between them as well as the corresponding

interface pressure and hoop stresses are Jess than for the case of equal temperatures. Hence, the

preceding analysis is conservative.

•

If the inner shell is hotter than the lead shell, an analysis is required. As described in the Lead Pour

Procedures in Section 8.4, the maximum allowed temperature differential measured between the inside

surface of inner shell and the outside surface of outer shell is 100°F (Section 8.4.2) and 160°F (Section

8.4.3) for the alternate lead pour procedure. Considering the bounding temperature differential of 160°F

for the inside surface of inner shell and the outside surface of outer shell, the temperature differential

between inner shell and lead shell is calculated to be less than J 00°F. Conservatively using a 100°F

temperature differential for the inner shell and the lead shell by assuming the temperature of the inner

shell to be 170°F and that of the lead to be 70°F, the inner radius of the stress-free lead shell at

70°F is 36.853 inches ( R;e ); the outer radius of the inner shell at 170°F is:

R =37.0 [1 +(8.71 x 10-6)(100)]

= 37.032 in

2.6.11.2-7

NAC-STC SAR


June 2018

Revision l 8B

The interference between the inner shell and the lead is 37.032 - 36.853 = 0.179 inch. To fully

accommodate this interference, the lead has to undergo a strain of 0.179/36.853 = 0.0049 inch/inch

or 0.49 percent. From Figure 24 ofNUREG/CR-048 I, the hoop stress in the lead is approximately

810 psi for a 0.0049 inch/inch strain. The interface pressure is:

q = (810) [( 40.5405)2

- (36.853)2

]

(40.5405) 2 + (36.853) 2

= 77 psi

The hoop stress in the inner shell becomes:

[(37)

2 + (35.5)2 l

Shis = (-77) (37)2 - (35.5)2

= -1862 psi

Note that the thermal expansion or contraction of a shell subjected to a constant pressure does not

affect the hoop stress; i.e.,

Sh = [(ka)2 + (kb)2 ]= [a

2

+ b2

]

q (ka)2 - (kb)2 q a 2 - b2

where

k = 1 + a ~T

This -1862 psi hoop stress (the inner shell is assumed to be 100°F hotter than the lead shell) reduces

to the previously calculated hoop stress of -1837 psi as both the inner shell and lead reach an

ambient temperature of 70°F. This does not take into account the beneficial effect of the creep

properties of the lead.

The axial stress in the inner shell also increases when the inner shell is 100°F hotter than the lead

shell. The axial stress of -2180 psi calculated when both the inner shell and lead shell are at 70°F

is recalculated for the inner shell temperature of l 70°F, a= 8.71 x 10-6 inch/inch/°F (Type 304

stainless steel):

2.6.11.2-8

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NAC-STC SAR

Docket No. 71-9235, Revision l7B

Se = (20.38 - 9.56)(10-6)(620 - 70) + (8.71 X 10-6)(] 70 - 70)

= 0.00595 in/in or 0.595 percent

June 2018

Revision l 8B

Referring to Figure 24 of NUREG/CR-048 I, the axial stress in the lead is approximately 820 psi.

The corresponding axial stress in the inner shell is -2 I 67 psi. As before, cooling of the inner shell

reduces this stress. The previous assumptions apply in arriving at this inner shell compressive

stress.

Temperature differentials between the inner and outer shells are of no consequence, since the axial

restraint between them is welded in place after cooldown, when the cask is at a uniform ambient

temperature. Welding of the outer shell and the bottom inner forging to the bottom outer forging

after cooldown is, therefore, a necessary fabrication step.

The question of buckling of the inner shell due to the combined effect of external pressure and

fabrication inaccuracies must also be addressed. According to the "ASME Boiler and Pressure

Vessel Code," Article NE-4221. I, the difference between the maximum and minimum inside

diameters at any cross section shall not exceed I percent of the nominal diameter at the cross

section under consideration. This amounts to (0.01)(71.0) = 0.71 inch. The relation between the

initial radial deviation, ro1, and the maximum and minimum diameter (Timoshenko and Gere,

Figure 7-10) is:

Dmax = Dnom + 2ro I

Dmin = Dnom - 2ro 1

thus

Dmax - Dmin = 4ro I

or

~D = 4ro1

Hence, the maximum initial radial deviation allowed is:

ffimax = ~D/4 = 0.71/4 = 0.1775 in

2.6. I 1.2-9

NAC-STC SAR


From Timoshenko and Gere, equation (7-15), page 2 93:

where

= 12,856 psi

E =27.76xl06 psiat170°F

V = 0.275

h = shell thickness= 1.50 in

R = shell radius= 36.25 in

Then from Timoshenko and Gere, equation (7-12), page 289:

qoc ~ [ 4 (I~ v 2)J[;r = Ser [: J

= 532 psi

June 2018

Revision l 8B

When the cylinder has fabrication inaccuracies, the external pressure, qYr, required to produce

yielding in the extreme fibers can be solved in the following equation (Timoshenko and Gere,

equation (e), page 296):

where

2 [Syr J Syr q YP - --;;;- + (I + 6mn) qcr g YP + --;;;-qcr = 0

Syp= 26,150 psi at l 70°F for Type 304 stainless steel

m = R/h = 36.25/1 .50 == 24.1667

n = mi/R = mi/36.25

2.6.11.2-10

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NAC-STC SAR


then

qvP2 - [l 082 + (I + 4m1)(532)] qvP + 575,660 = 0

June 2018

Revision 18B

The value of m1 can vary from 0.0 inches (perfect cylinder) up to 0.1775 inch (maximum allowed

according to the "ASME Boiler and Pressure Vessel Code"). Solving qyp for varying values of m1

gives the following:

Initial Radial Yield

Deviation Pressure

m1 (in) gyp (psi)

0.001 530

0.06 443

0.12 389

0.1775 351

Thus, the margin of safety against yielding for the inner shell with maximum allowed radial

deviation subjected to 77 psi lead pressure (inner shell temperature is assumed to be 100°F higher

than lead temperature) is:

351 MS = - - 1 = +3.55

77

Since the margin of safety for this conservative load case is positive, the inner shell does not buckle

when subjected to the lead pressure produced during the cooling of the cask .

2.6.11.2-11

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Atlanta Corporate Headquarters: 3950 East JOfleS Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

June 2018

Revision 18B

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List of Effective Pages

Chapter 1 Page 2.4-1 ..................................... Revision 18

Page 2.4.1-1 .................................. Revision 18

Page 1-i thru 1-iv .......................... Revision 18 Page 2.4.2-1 .................................. Revision 18

Page 1-v ...................................... Revision 18B Page 2.4.3-1 .................................. Revision 18

Page 1-vi thru I-ix ........................ Revision 18 Pages 2.4.4-1 thru 2.4.4-6 ............. Revision 18

Pages 1-1 thru 1-12 ....................... Revision 18 Pages 2.4.4-7 thru 2.4.4-8 ........... Revision 18B


Page 1.1-4 ................................... Revision 18B Pages 2.4.4-10 ............................. Revision 18B



Page 1.2-9 ................................... Revision 18B Pages 2.5.1-1 thru 2.5.1-3 8 ........... Revision 18

Pages 1.2-10 thru 1.2-42 ............... Revision 18 Pages 2.5.2-1 thru 2.5.2-29 ........... Revision 18

Pages 1.2-43 thru 1.2-44 ............. Revision 18B Pages 2.6-1 thru 2.6-2 ................... Revision 18

Pages 1.2-45 thru 1.2-49 ............... Revision 1 8 Pages 2.6.1-1 thru 2.6.1-7 ............. Revision 18

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13 drawings in Sections

2.10.6.6 and 2.10.6.7


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Chapter 6

Chapter 5

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Chapter 7

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NAC-STC SAR


Table of Contents

June 2018

Revision 18B

3.0 THERMAL EVALUATION .......................................................................................... 3.1-l

3.1 Discussion ......................................................................................................................... 3.1-1

3.1.1 Directly Loaded (Uncanistered) Fuel. .............................................................. 3.1-'3

3.1.2 Canistered Yankee Class Fuel ......................................................................... 3.1-4

3.1.3 Canistered Connecticut Yankee Fuel ............................................................... 3.1-6

3.1.4 Canistered Greater Than Class C Waste .......................................................... 3.1-8

3.1.5 Directly Loaded (Uncanistered) PWR High Burnup Fuel ............................... 3.1.9

3.2 Summary of Thermal Properties of Materials .................................................................. 3.2-l

3.2.1 Conductive Properties ...................................................................................... 3.2-1

3.2.2 Radiative Properties ......................................................................................... 3.2-1

3 .2 .3 Convective Properties ...................................................................................... 3 .2-7

3.2.4 Neutron Shield (NS-4-FR) Thermal Conductivity .......................................... 3.2-8

3.3 Technical Specifications for Components ........................................................................ 3.3-l

3.3.1 Radiation Protection Components ................................................................... 3.3-1

3.3.2 Safe Operating Ranges ..................................................................................... 3.3-2

3.4 Thermal Evaluation for Normal Conditions of Transport ................................................ 3.4-1

3.4.1 Thermal Models ............................................................................................... 3.4-1

3.4.2 Maximum Temperatures ................................................................................ 3.4-26

3.4.3 Minimum Temperatures ................................................................................. 3.4-30

3.4.4 Maximum Internal Pressure ........................................................................... 3.4-30

3.4.5 Maximum Thermal Stresses .......................................................................... 3.4-44

3.4.6 Summary ofNAC-STC Performance for Normal Conditions of Transport .. 3.4-44

3.4.7 Normal Heat-up Transient ............................................................................. 3.4-44

3-i

NAC-STC SAR

Docket No. 71-9235

Table of Contents

(Continued)

March 2017

Revision 18

3.5 Hypothetical Accident Thermal Evaluation ...................................................................... 3.5-l

3.5.1 Thermal Model. ................................................................................................ 3.5-I

3.5.2 Package Conditions and Environment.. ........................................................... 3.5-4

3.5.3 Package Temperatures ..................................................................................... 3.5-5

3.5.4 Maximum Internal Pressure ............................................................................. 3.5-7

3.5.5 Maximum Thermal Stresses .......................................................................... 3.5-10

3.5.6 Evaluation of Package Performance for Hypothetical Accident

Thermal Conditions ....................................................................................... 3.5-10

3.6 Thermal Evaluation - STC-LACBWR ......................................................................... 3.6-1

3.6.1 Discussion-STC-LACBWR .............................................................................. 3.6.1-1

3.6.2 Summary of Thermal Properties of Materials- STC-LACBWR ...................... 3.6.2-1

3.6.3 Technical Specifications for Components - STC-LACBWR ....................... 3.6.3-1

•

3 .6.3 .1 Radiation Protection Components .................................................. 3 .6.3-1

3.6.3.2 Safe Operating Ranges .................................................................... 3.6.3-2 •

3.6.4 Thermal Evaluation for Normal Conditions ofTranspo11- STC-LACBWR .. 3.6.4-1

3 .6.4.1 Thermal Models .............................................................................. 3 .6.4-1

3 .6.4.2 Maximum Temperatures ................................................................. 3 .6.4-7

3.6.4.3 Minimum Temperatures .................................................................. 3.6.4-8

3.6.4.4 Maximum Internal Pressure ............................................................ 3.6.4-8

3.6.4.5 Maximum Thermal Stresses ......................................................... 3.6.4-13

3.6.4.6 Summary of STC-LACBWR Performance for Normal Conditions

of Transport. .................................................................................. 3 .6 .4-13

3.6.5 Hypothetical Accident Thermal Evaluation - STC-LACBWR ................... 3.6.5-1

• 3-ii

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•

NAC-STC SAR

Docket No. 71-9235

Table of Contents

(Continued)

March 2017

Revision 18

3.7 Thermal Evaluation- STC-WVDP .............................................................................. 3.7-1

3.7.1 Discussion - STC-WVDP HL W Overpack and Contents ................................ 3.7.1-1

3.7.2 Summary ofThermal Properties ofMaterials-STC- WVDP ......................... 3.7.2-1

3.7.3 Technical Specifications for Components- STC - WVDP ........................... 3.7.3-1

3.7.3.1 Radiation Protection Components .................................................. 3.7.3-1

3.7.3.2 Safe Operating Ranges .................................................................... 3.7.3-1

3.7.4 Thermal Evaluation for Normal Conditions ofTranspo11-MPC - WVDP .... 3.7.4-1

3.7.4.1 Thermal Model.. .............................................................................. 3.7.4-1

3.7.4.2 Maximum Temperatures ................................................................. 3.7.4-3


3. 7.4.4 Maximum Internal Pressure ............................................................ 3. 7.4-4

3.7.4.5 Maximum Thermal Stresses ........................................................... 3.7.4-4

3.7.4.6 Summary of STC- WVDP Performance for Normal Conditions

of Transport ..................................................................................... 3. 7 .4-4

3. 7 .5 Hypothetical Accident Thermal Eva! uation - STC- WVDP ......................... 3. 7 .5-1

3.8 Thermal Evaluation - STC-High Burnup Directly Loaded Fuel (STC-HBU) ............. 3.8-1

3.8.1 Discussion - STC-HBU and Contents ........................................................... 3.8.1-1

3.8.2 Summary of Thermal Properties of Materials - STC-HBU .......................... 3.8.2-1

3.8.3 Technical Specifications for Components - the STC-HBU .......................... 3.8.3-1

3.8.3.1 Radiation Protection Components .................................................. 3.8.3-1

3.8.3.2 Safe Operating Ranges .................................................................... 3.8.3-2

3.8.4 Thermal Evaluation for Normal Conditions of Transport - the STC-HBU .. 3.8.4-1

3.8.4.1 Thermal Models .............................................................................. 3.8.4-l

3.8.4.2 Maximum Temperatures ................................................................. 3.8.4-6


3.8.4.4 Maximum Thermal Stresses ........................................................... 3.8.4-8

3.8.4.5 Summary of the STC-HBU Performance for Normal Conditions of

Transpo1i ......................................................................................... 3.8.4-8

3.8.5 Hypothetical Accident Thermal Evaluation - the STC-HBU ........................ 3.8.5-1

3.8.6 Maximum Pressure During Normal and Hypothetical Accident Conditions (HAC)

of Transport .................................................................................................... 3.8.6-1

3-iii

NAC-STC SAR


List of Figures

Figure 3.1-1 Definition of the Gap between the Basket and the Inner Shell

June2018

Revision 18B

for the Horizontal Position of the Cask ........................................................... 3 .1-10

Figure 3.1-2 Definition ofthe Gap between the Yankee-MPC Basket, Canister,

and the Inner Shell for the Horizontal Position of the NAC-STC .................. 3 .1-10

Figure 3.1.3 Basket Orientation and Gap between the CY-MPC Basket, Canister

and the Inner Shell for the Horizontal Position of the NAC-STC .................. 3 .1-11

Figure 3.2-1 Radial Temperature Profile versus NS-4-FR Thermal Conductivity

for Directly Loaded Fuel.. ................................................................................. 3.2-9

Figure 3.3-1 NS-4-FR Developer's Test Results Letter. ........................................................ 3.3-5

Figure 3.3-2 JAPC NS-4-FR Technical Data ........................................................................ 3.3-6

Figure 3.4-1 Three-Dimensional ANSYS Model for Directly Loaded Fuel ....................... 3.4-46

Figure 3.4-2 Design Basis Directly Loaded PWR Fuel Assembly

Axial Flux Distribution ................................................................................... 3.4-47

Figure 3.4-3 Horizontal View of the ANSYS Model for Directly Loaded Fuel

Containing the Support Disk, Fuel Assembly Elements and Shell. ................ 3.4-48

•

Figure 3.4-4 Detailed View of a Portion of the ANSYS Directly Loaded •

Fuel Basket Model .......................................................................................... 3 .4-49

Figure 3.4-5 Isometric View of the Directly Loaded Fuel Elements

for the Thermal Model .................................................................................... 3.4-50

Figure 3.4-6 Isometric View of the 180-Degree Section Cask Thermal

Model for Directly Loaded Fuel ..................................................................... 3.4-51

Figure 3.4-7 Detailed View of Basket and Shells of the 180-Degree

Section Cask Thermal Model for Directly Loaded Fuel.. ............................... 3.4-52

Figure 3.4-8 Plan View of the Directly Loaded Fuel 180-Degree Section

Cask Thern1al Model ....................................................................................... 3.4-53

Figure 3.4-9 Directly Loaded Fuel Assembly Thermal Model ........................................... 3.4-54

Figure 3 .4-10 Detailed View of a Single Fuel Rod in the Directly Loaded Fuel

Assembly Thermal Model ............................................................................... 3.4-55

Figure 3.4-11 Directly Loaded Fuel Basket Temperature Distribution for the Steel

Support Disk with Helium .............................................................................. 3.4-56

Figure 3.4-12 Directly Loaded Fuel Basket Temperature Distribution for the

Aluminum Heat Transfer Disk with Helium .................................................. 3.4-57

• 3-iv

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NAC-STC SAR June 2018

Revision l 8B Docket No. 71-9235, Revision l 7B

Figure 3.4-13

Figure 3.4-14

Figure 3.4-15

Figure 3 .4-16

Figure 3.4-17

Figure 3 .4-18

Figure 3 .4-19

Figure 3.4-20

Figure 3.4-21

Figure 3.4-22

Figure 3.4-23

Figure 3.4-24

Figure 3 .4-25

Figure 3.4-26

Figure 3.4-27

Figure 3.4-28

Figure 3.4-29

Figure 3.4-30

Figure 3.4-31

Figure 3.4-32

Figure 3.4-33

Figure 3.4-34

List of Figures

(Continued)

Directly Loaded Fuel Basket Temperature Distribution for the Steel

Support Disk with Air. .................................................................................... 3.4-58

Directly Loaded Fuel Basket Temperature Distribution for the

Aluminum Heat Transfer Disk with Air ......................................................... 3.4-59

Isometric View of Quarter Symmetry Heat-up Transient

Model for Directly Loaded Fuel ..................................................................... 3 .4-60

Heat-up Transient Thermal Response of the Directly Loaded

Basket Aluminum Disk ................................................................................... 3.4-61

Heat-up Transient Average Temperature Response for Directly

Loaded Fuel Basket Aluminum Disk and Inner Shell Wall ........................... 3.4-62

Heat-up Transient Thermal Response for the Directly

Loaded Fuel Basket Steel Support Disk ......................................................... 3 .4-63

Heat-up Transient Average Temperature Response for the Directly

Loaded Fuel Basket Steel Support Disk and Inner Shell Wall ....................... 3.4-64

Three-Dimensional ANSYS Model for Yankee-MPC Canistered Fuel ......... 3.4-65

Design Basis Yankee Class Canistered Fuel Assembly Axial

Power Distribution .......................................................................................... 3.4-66

Fuel Assembly Model for Yankee-MPC Canistered Fuel .............................. 3.4-67

Fuel Tube Model for Yankee-MPC Canistered Fuel ...................................... 3.4-68

Two-Dimensional Yankee Reconfigured Fuel Assembly Model ................... 3.4-69

Yankee Damaged Fuel Locations in the Three-Dimensional Cask Model... .. 3.4-70

Three-Dimensional Cask Thermal Model for the CY-MPC ........................... 3.4-71

Three-Dimensional Cask Thermal Model for CY-MPC - Cross-Section ...... 3.4-72

Design Basis Connecticut Yankee Fuel Assembly Axial

Power Distribution .......................................................................................... 3 .4-73

Quarter-Symmetry Connecticut Yankee Fuel Assembly Model .................... 3.4-74

Fuel Tube Model for Connecticut Yankee Canistered Fuel ........................... 3.4-75

CY-MPC GTCC Transport Configuration Finite Element Model ................. 3 .4-76

CY-MPC GTCC Thermal Model Cross-Section ............................................ 3.4-77

Personnel Barrier Thermal Model .................................................................. 3 .4-78

Temperature Results for the Personnel Barrier. .............................................. 3 .4-79

3-v

NAC-STC SAR

Docket No. 71-9235

List of Figures

(Continued)

Figure 3.5-1 NAC-STC Hypothetical Accident Conditions ANSYS Model

March 2017

Revision 18

for Directly Loaded Fuel.. ............................................................................... 3 .5-11

Figure 3.5-2 NAC-STC Hypothetical Accident Conditions Temperature History

for the Directly Loaded Basket ....................................................................... 3.5-12

Figure 3.5-3 NAC-STC Hypothetical Accident Conditions Temperature History

for CY-MPC Fuel ........................................................................................... 3.5-15

Figure 3.6-1 Three-Dimensional ANSYS Model for STC-LACBWR ........................... 3.6.4-14

Figure 3.6-2 Three-Dimensional Cask Thermal Model for STC-LACBWR-

Cross-Section ............................................................................................... 3.6.4-15

Figure 3.6-3 Damaged Fuel Locations in the Three-Dimensional Cask and

Canister Model (Cross-Section) ................................................................... 3.6.4-16

Figure 3.6-4 Design Basis LACBWR Fuel Assembly Axial Power Distribution ............ 3.6.4-17

Figure 3.6-5 Fuel Assembly Model for LACBWR Fuel .................................................. 3.6.4-18

Figure 3.6-6 Two-Dimensional MPC-LACBWR Fuel Tube Model

(Standard Fuel Tube with BORAL Plate) .................................................... 3.6.4-19


(Standard Fuel Tube without BORAL Plate) ............................................... 3.6.4-20


(Fuel Tube in the Slots Containing DFC with BORAL) ............................. 3.6.4-21


(Fuel Tube in the Slots Containing DFC without BORAL) ........................ 3.6.4-22

Figure 3.7-1 Three-Dimensional ANSYS Model for STC-WVDP ................................... 3.7.4-6

Figure 3.7-2 Three-Dimensional Model for STC- WVDP- Cross-Section ...................... 3.7.4-7

Figure 3.8-1 Configuration Definition for the STC-HBU ................................................ 3.8.4-10

Figure 3.8-2 Three-Dimensional ANSYS Model for the STC-HBU ............................... 3.8.4-11

Figure 3.8-3 Three-Dimensional Cask Thermal Model for the STC-HBU - Cross-Section

(Configuration 8) ......................................................................................... 3.8.4-12

Figure 3.8-4 HBU Fuel Assembly Axial Power Distribution ........................................... 3.8.4-13

Figure 3.8-5 Fuel Assembly Model for the STC-HBU Fuel ............................................ 3 .8.4-14

Figure 3.8-6 Two-Dimensional Fuel Tube Model for the STC-HBU .............................. 3.8.4-15

3-vi

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Revision 18B Docket No. 71-9235, Revision 17B

List of Tables

Table 3.1-1 Thermal Analysis Bounding Conditions - Normal Transport

Conditions ....................................................................................................... 3 .1-12

Table 3.2-1 Thermal Properties of Solid Neutron Shield (NS-4-FR) ................................ 3.2-10

Table 3.2-2 Thermal Prope1iies of Stainless Steel ............................................................. 3 .2-10

Table 3.2-3 Thermal Prope1iies of Chemical Copper Lead ............................................... 3 .2-11

Table 3.2-4 Thermal Properties of Type 606 l-T6 and 606 l-T65 I Aluminum

Alloy ............................................................................................................... 3.2-11

Table 3.2-5 Thermal Prope1iies of Helium ........................................................................ 3.2-12

Table 3.2-6 Thermal Prope1iies of Dry Air. ....................................................................... 3 .2-12

Table 3.2-7 Thermal Properties of Copper ......................................................................... 3 .2-12

Table 3.2-8 Thermal Properties of B4C .............................................................................. 3 .2-13

Table 3.2-9 Thermal Properties of Zircaloy and Zircaloy-4 Cladding ............................... 3 .2-13

Table 3.2-10 Thermal Properties of Fuel (U02) ................................................................... 3 .2-13

Table 3.2-11 Thermal Properties ofBORAL and Talbor Sheet... ........................................ 3.2-14

Table 3.4-1 Maximum Component Temperatures - Normal Transport

Conditions, Maximum Decay Heat and Maximum Ambient

Temperature - Directly Loaded and Canistered Configurations ..................... 3 .4-80

Table 3.4-2 Maximum Component Temperatures - Normal Transport

Conditions, Maximum Decay Heat, Minimum Ambient

Temperature - Directly Loaded and Canistered Configurations ..................... 3.4-81

Table 3.4-3 Maximum Component Temperatures - Normal Transport Conditions,

Maximum Decay Heat, Low Ambient, for Directly Loaded Fuel .................. 3.4-82

Table 3.4-4 NAC-STC Thermal Performance Summary for Normal

Conditions of Transport .................................................................................. 3 .4-83

Table 3.4-5 Maximum Cask Component Temperatures in Normal Conditions

of Transport ..................................................................................................... 3 .4-84

Table 3.4-6 Maximum Component Temperatures for Yankee-MPC Damaged Fuel ........ 3.4-85

Table 3.4-7 Maximum Component Temperatures for CY-MPC Damaged Fuel.. ............. 3.4-85

Table 3.4-8 Westinghouse 15 X 15 Fuel Assembly Characteristics .................................. 3.4-85

Table 3.4-9 Directly Loaded Fuel Basket Component Volumes ....................................... 3.4-86

Table 3.5-1 Maximum Component Temperatures - Hypothetical Accident

Conditions Fire Transient ............................................................................... 3 .5-16

Table 3.6-1 Thermal Analysis Bounding Conditions - Normal Transport

Conditions ...................................................................................................... 3 .6.1-4

3-vii

NAC-STC SAR

Docket No. 71-9235

List of Tables (continued)

March 2017

Revision 18

Table 3.6-2 Thermal Properties of Helium ....................................................................... 3.6.2-2

Table 3.6-3 Gaps within the STC-LACBWR Three-Dimensional Thermal Model ......... 3.6.2-3


Maximum Decay Heat and Maximum Ambient Temperature -

STC-LACBWR ............................................................................................ 3.6.4-23


Maximum Decay Heat, Minimum Ambient Temperature -

STC-LACBWR ............................................................................................ 3.6.4-24

Table 3.6-6 Maximum Temperature of the Fuel, Basket, and Canister- Hypothetical

Table 3.7-1

Table 3.7-2

Table 3.7-3

Table 3.7-4

Accident Condition Fire Transient.. ............................................................... 3.6.5-3

Thermal Analysis Bounding Conditions - Normal Transport

Conditions ...................................................................................................... 3.7.1-3

Thermal Properties of Glass ........................................................................... 3. 7 .2-2

Gaps within the STC-WVDP Three-Dimensional Thermal Model.. ............. 3.7.2-2

Maximum Component Temperatures - Normal Transport Conditions,

•

Maximum Decay Heat and Maximum Ambient Temperature - •

STC-WVDP ................................................................................................... 3.7.4-8


Maximum Decay Heat, Minimum Ambient Temperature -

STC- WVDP .................................................................................................. 3.7.4-9

Table 3.7-6 Maximum Temperature of the HLW and Contents, Basket, and HLW

Overpack - Hypothetical Accident Condition Fire Transient ......................... 3.7.5-2

Table 3.8-1 Thermal Analysis Bounding Conditions - Normal Transpo11 Conditions .... 3.8.1-4

Table 3.8-2 Minimum Thermal Conductivities of Neutron Absorber (MMC, 45wt%

B4C) ............................................................................................................... 3.8.2-2

Table 3.8-3 Gaps within the STC-HBU Three-Dimensional Thermal Model .................. 3.8.2-3

Table 3.8-4 Maximum Component Temperatures-Normal Transport Conditions,

Maximum Decay Heat, Maximum Ambient Temperature, among Three

Configurations - the STC-HBU ................................................................... 3 .8.4-16


Maximum Decay Heat, Minimum Ambient Temperature, among Three

Configurations - the STC-HBU ................................................................... 3.8.4-17

Table 3.8-6 Maximum Temperature of the STC-HBU - Hypothetical Fire Accident

Condition ........................................................................................................ 3 .8.5-2 •

3-viii

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NAC-STC SAR


June 2018

Revision 18B

analysis of the cask heat-up condition has been performed using the ANSYS finite element

program. The model, Figure 3.4-15, represents a quarter symmetry slice from the center section

of the cask and includes the fuel assemblies, fuel tubes, steel support disks, aluminum heat

transfer disks, and cask body wall. Each of these areas and components are modeled using the

material properties and detail represented in heat transfer finite element models discussed earlier,

with added heat transfer enhancement representing radiation between the tubes, and between the

tubes and the cask inner shell in the spaces between the support disks and the aluminum heat

transfer disks. In order to capture the influence of the initial vacuum drying process, the boundary

conditions starting the transient represented all components at 70°F, fuel assembly design basis

heat load of 0.85 kilowatts, and the cavity evacuated. At twenty-four hours into the transient,

helium was added to the model representing the normal operating procedure of back filling the

cavity with helium following completion of the drying process.

Component temperature profiles were obtained for each time step through cask steady state

conditions. Figures 3.4-16, 3.4-17, 3.4-18, and 3.4-19 present the transient temperature results

for the aluminum heat transfer disk; aluminum heat transfer disk average temperature and average

inner shell temperature; support disk; and support disk average disk temperature and average

inner shell temperature, respectively.

It is concluded from these results that a steady state heat flow is established throughout the cask

at approximately I 00 hours after fuel load with actual peak temperatures reached at about 240

hours after fuel load. Temperatures from this analysis are used as input loading to evaluate the

potential for basket and cask wall interference resulting from thermal expansion.

For canistered fuel, the canister configuration has been evaluated to ensure that the canister at the

steady state hot condition can be installed in an NAC-STC at the steady state cold condition. The

cold condition temperature is limited to 0°F, since this is the limiting temperature for operation

of the transfer cask. The transfer cask is used to install the canister in the NAC-STC .

3.4-45

NAC-STC SAR


Figure 3.4-1 Three-Dimensional ANSYS Model for Directly Loaded Fuel

3.4-46

June 2018

Revision l 8B •

•

•

•

•

•

NAC-STC SAR


June 2018

Revision l 8B

Figure 3.4-2 Design Basis Directly Loaded PWR Fuel Assembly Axial Flux Distribution

z 0

8 IX ...)

w ::> L&.

~ .::. u ,c(

u.. 0 I-X 0 w ::x::

,oo,;

75~

50%

25~

o"

(100~ •. 53)

•·

(0%..53)

0 . 1 .2 .3 .4 .5 .6 . 7 .8 .9 1.0 1.1

RELA 11VE ?OWER

3.4-47

(80%.1.1)

(15%, 1.1)

NAC-STC SAR

Docket No. 71-9235, Revision l7B June 2018

Revision l 8B

Figure 3.4-3 Horizontal View of the ANSYS Model for Directly Loaded Fuel Containing the

Support Disk, Fuel Assembly Elements and Shell

3.4-48

Fuel

Basket

Basket/Inner Shell Gap

Cask Inner Shell

Lead Ga11111a Shield

Cask Outer Shell

Radial Neutron Shield

Neutron Shield Shell

•

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•

•

•

•

NAC-STC SAR


June2018

Revision 18B

Figure 3 .4-4 Detailed View of a Portion of the AN SYS Directly Loaded Fuel Basket Model

Fuel

Composite Fuel Tube (stainless steel, BORAL

Gap between Fuel & Fuel Tube & Air Gaps)

3.4-49

ii

Basket

NAC-STC SAR


June2018

Revision l 8B

Figure 3.4-5 Isometric View of the Directly Loaded Fuel Elements for the Thermal Model

3.4-50

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NAC-STC SAR


June2018

Revision l 8B

Figure 3.4-6 Isometric View of the 180-Degree Section Cask Thermal Model for Directly

Loaded Fuel

3.4-51

NAC-STC SAR


June2018

Revision 18B

Figure 3.4-7 Detailed View of Basket and Shells of the 180-Degree Section Cask Thermal

Model for Directly Loaded Fuel

Stainless Steel Support Disk

Aluminum Heat Transfer Fin

Basket/Inner Shell Gap

Inner Shell

Ganna Shield

Gap

Outer Shell


Radial Neutron Shield Shell

3.4-52

•

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•

•

NAC-STC SAR


June 2018

Revision 18B

Figure 3.4-8 Plan View of the Directly Loaded Fuel 180-Degree Section Cask Thermal Model

-------Tl--- Maximum Gap

3.4-53

( 0 .13 inches)

Basket

Basket Inner Shell Contact Surface

Inner Shell

NAC-STC SAR


Figure 3.4-9 Directly Loaded Fuel Assembly Thermal Model

Plane of S}'ITl11etry

Exterior of Fuel Assembly Model

3.4-54

June 2018

Revision 18B

Center of Fuel Assembly

Plane of Syn111etry

•

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•

NAC-STC SAR


June 2018

Revision 18B

Figure 3.4-10 Detailed View of a Single Fuel Rod in the Directly Loaded Fuel Assembly

Thermal Model

Fuel Cladding Cavity Gas

3.4-55

NAC-STC SAR


June2018

Revision 18B

Figure 3.4-11 Directly Loaded Fuel Basket Temperature Distribution for the Steel Support

Disk with Helium

SteeJ 01sk

Transport/Hel1um

Maximum Gap ~-_,...20

7 93

.:10 EJD68 38 EJ EJ '----79~ 95

EJEJEJ6~ 23

/ "

79 87

Region of Contact'-- 01

~ (Temperatures in °F)

3.4-56

•

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•

•

•

NAC-STC SAR


June 2018

Revision 18B

Figure 3.4-12 Directly Loaded Fuel Basket Temperature Distribution for the Aluminum Heat

Transfer Disk with Helium

35 39

Aluminum Fin Region

Transport/Helium

(Temperatures in °F)

Maximum Gap __ ___,....16

B 85

04 46 60

29 72 87

01 43 57

EJ .___7 2___...__.__..

80

11

3.4-57

NAC-STC SAR


June 2018

Revision l 8B

Figure 3.4-13 Directly Loaded Fuel Basket Temperature Distribution for the Steel Support

Disk with Air

Maximum Gap ____ ,28

93

10 55 70

51 52 37 81 97

52 67

80 89

Steel Disk Transport/AIR

Region of Contact

~ 03


3.4-58

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•

NAC-STC SAR


June 2018

Revision 18B

Figure 3.4-14 Directly Loaded Fuel Basket Temperature Distribution for the Aluminum

Heat Transfer Disk with Air

Aluminum Fin Transport/AIR

Maximum Gap

06

32

Reg1on of Contact


3.4-59

__ _.,..16

86

63

75 90

60

74 82

NAC-STC SAR


June2018

Revision l 8B

Figure 3.4-15 Isometric View of the Quaiter Symmetry Heat-up Transient Model for

Directly Loaded Fuel

3.4-60

•

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•

•

NAC-STC SAR


June2018

Revision I SB

Figure 3.4-16 Heat-up Transient Thermal Response of the Directly Loaded Basket

Aluminum Disk

TEMP 800

720

640

560

.:180

.:100

320

2.ao

160

80

~~

~ -;:::::------

If ~ ~ ~ ~ l ---

~ 'l ~ ~

r

' fl 0

0 80 160 40 120

3.4-61

240 200 280

.1 ~ ~L Q ~ - -

158 AL .1.1

1 F;] f. I 15 ~ _::;i AL .1 ::l

:, 34 AL 23

::, li3~~ §i

320 360

HRS

400

NAC-STC SAR


June2018

Revision 18B

Figure 3.4-17 Heat-up Transient Average Temperature Response for Directly Loaded Fuel

Basket Aluminum Disk and Inner Shell Wall

TEMP 500

A50 ~

Ll DO /

350 I

I 300

250

200

150

100

I ~ v-

I V

/ I I V

II V

50

0 0 80 150

40 120

3.4-62

240 200 280

l'.'."[NS-T AV

TSHL-T AV

320 360

4

HRS

00

•

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•

•

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•

NAC-STC SAR


June 2018

Revision 18B

Figure 3.4-18 Heat-up Transient Thermal Response for the Directly Loaded Fuel Basket

Steel Suppo1i Disk

TEMP 800

720

640

1 ~ ~~ 0 560

.::180

400

320

2~0

160

80

r::::::-- 170 s~ 1 1

~ ~ 169 ss 15

IP, ~ [:::::::= u--::r --;:J..; r-::J

') g5 55 23 ---~ / J ? 95 s~ 28 - 1q335~ 32

~ v/% ~ ...... -::i 37 S= .35 -~

y ~

' ~ y 0 HRS

0 80 160 240 320 400 40 120 200 280 360

3.4-63



Figure 3.4-19 Heat-up Transient Average Temperature Response for the Directly Loaded

Fuel Basket Steel Supp011 Disk and Inner Shell Wall

TEMP 500

n1SK-T AV Ll50 ----LIDO

/ /

350

300

250

200

150

100

I I

I T SHL-T AV

~ ~

I /

/ I V

I

II V

50

0 HRS 0 80 160 240 320 ADO

-40 120 200 280 360

3.4-64

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•

NAC-STC SAR


June2018

Revision l 8B

Figure 3.4-20 Three-Dimensional ANSYS Model for Yankee-MPC Canistered Fuel

Spacer

Canister Bottom Plate

3.4-65

Gap


Spacer

Outer Shell

Lead

NAC-STC SAR

Docket No. 71-9235, Revision 17B June 2018

Revision 18B

Figure 3.4-21 Design Basis Yankee Class Canistered Fuel Assembly Axial Power Distribution

L

0

+:>

l .

U 1.0 d

LL

C

0

.9

.8

2

8

6

4

I

I \ r1 \

7 \ .,....( .7

L

~

(/")

z7 \ 0

'a.,

3: 0

0...

a; ::,

LL

.4

.J

.2

. l

6 /

8

6

4

2

0 0 .2 .4 .6 8

. l .J .5 .7 .9

Fraction of Active Fuel Length

3.4-66

\ \

1

•

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•

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•

•

NAC-STC SAR


Figure 3.4-22 Fuel Assembly Model for Yankee-MPC Canistered Fuel

3.4-67

June 2018

Revision 18B

NAC-STC SAR


June20]8

Revision l SB

Figure 3.4-23 Fuel Tube Model for Yankee-MPC Canistered Fuel

Fuel Tube

Helium Gap

Helium Gap

Stainless Steel Clad for BORAL Plate

Helium Gap

Aluminum Clad ofBORAL Plate

Core Matrix ofBORAL Plate

Aluminum Clad ofBORAL Plate

3.4-68

•

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•

NAC-STC SAR

• Docket No. 71-9235, Revision l 7B

Figure 3.4-24 Two-Dimensional Yankee Reconfigured Fuel Assembly Model

Shell Casing

1Desi gn Case :rfa_tr

•

Fuel Rod Helium Tube

• 3.4-69

June 2018

Revision I SB

ANSYS APR 9 16:34:1 PLOT NO. ELEMENT TYPE

zv D1ST=2.1 XF YF

Shell Casing

Radiation Element (Typ.)

NAC-STC SAR

Docket No. 71-9235, Revision l7B June 2018

Revision l 8B

Figure 3.4-25 Yankee Damaged Fuel Locations in the Three-Dimensional Cask Model

Damaged fuel can is restricted to these positions

Debris of the 20 failed fuel rods is concentrated in the center 7 .1 inches of the active fuel region

3.4-70

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NAC-STC SA R

Docket o. 71-9235 , Rev ision 17B

June2018

Rev ision I 8B

Figure 3.4-26 Three- Dimensional Cask Thermal Model fo r the CY-MPC

Cask Outer Bottom

r-Canister Bottom Plate I I

Fuel Basket Bottom Weldment

Lower Neut ran Shield

Support Disks

Heat Transfer



Outer Shell

Gamma Shield

Inner Shell

Canister Shell

Fuel Basket Top Weldment

Note : canister fill gas , cask fill gas, and bottom spacer not shown for clariq.1 .

Cask Inner Lid

Upper Neutron Shield

Cask Outer Lid

3.4-7 1

Shield Lid

Lid

NAC-STC SAR

Docket No. 71-9235 , Revisi on 178

June20 18

Revi s ion I 8B

Figure 3.4-27 Three-Dimensiona l Cask Thermal Model for CY-MPC - Cross-Section

Oversize FUel Assembly

~ Neutron Shield Shell

Neutron Shield

Gamma Shield

Cask Inner Shell Fuel Tubes

Helium (Cask Gas)

Canister Shell

No Gaps

Support Disk/Heat Transfer Disk/Helium

3.4-72

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NAC-STC SAR

Docket No. 71-9235 , Rev ision l 7B

Ju ne 20 18

Rev ision I 8B

Figure 3.4-28 Des ign Basis Connecticut Yankee Fuel Assembl y Axial Power Distribution

1.2

1 . 1 :1. 1

I L I I I

0. , - -----. ---.-::, C: L. ::, al I

0.6 T

[_

Q)

> ... ra I

Q) 0.4

0 .4:33 a::: -,--

0.2

0

0 0.1 0.2 0 .3 0 .4 0.5 0.6 0 .7 0.8 0.9

Fraction of Core Height

3.4-73

NAC-STC SAR

Docket No. 71-9235 , Rev ision 17B

F igure 3.4-29 Quarter-Symmetry Connect icut Yankee Fue l Assembly Model

Quarter - Symmetry Bounda r y

Center of Fuel Assembly

Quarter-Symmetry Boundary

3.4-74

Inner Surface of Fuel Tube

Fuel Cladding

Helium

June 20 18

Revi sion l 8B •

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•

NAC-STC SAR

Docket No. 71-9235 , Revision I 7B

June 2018

Revision I 8B

Figure 3.4-30

Heat Flu:~

Fuel Tube Model for Connecticut Yankee Cani stered Fuel

y

Core Matri:-: of BORAL Plate

Fuel Tube

Gap r sca,o,ess

• • ~

Steel Clad for BORAL Plate

I •

Helium Gap

Inner Edge of Support/Heat

/ Transfer Disk • Slot

L,__ _ __..._,__•_,___ __ -r-'-•---- -----x

Aluminum Clad of BORAL Plate

Radiation Link Element 1

3.4-75

NAC-STC SAR

Docket No. 71-9235 , Rev ision I 7B

June 20 18

Revision I 8B

Figure 3.4-31 CY-MPC GTCC Transport Configuration Finite Element Model

Neut ron Shield Shell

Neutron Shield

Cask Outer Shell

0.01 5 Air GaR

GTCC Canister

Tube Array Weldment

Shield Shell Weldment

GTCC Waste

3.4-76

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NAC-STC SAR

Docket No. 71-9235 , Revi sion l 7B

Figure 3.4-32 CY-MPC GTCC Therma l Model Cross-Section

Ne ut ron Shield Shell

Neut ron Shield

Cask Outer Shell

Lead

Cask Inner

Helium Fi ll Gas

-

-

Shield Shell Weldme 111

Tube Array We ld,

GTCC Waste Helium

----11 en t

and --Gas

-Tube Array We ld mcnt

GTCC Waste Helium

and --Gas

Tube Array Weldm ent

GTCC Waste Helium

Tube Array Weld

and --Gas

ment

~ ·--- -----

-

-

-

-

-i--

--

3.4-77

/

.--

--

June 20 18

Revi sion 18B

0.015AirGap

GTCC Canister Shell

GTCC Shield Shell Support Disk

Y (Radia l)

z J

Tran sport Cask and Basket Center Line

NAC-STC SAR

Docket No. 71-9235, Revision I 7B

F igure 3.4-33 Personnel Barrier Therma l Model

l 12.8 in .

!

Cask Surface

Location of

the Personnel

Barrier

3.4-78

11111111

lllli ll\11

~

~

lli, '\

I 'J

~,___r-

"r-"1--1--

r-r-r-

' '

I'-" i'-1'-

I'-..._

I'-

1' i'-1'-

I'-

June 20 18

Revi sion 18B

', i'-,1'-

1--'I'-

1--, I'-"

', 'i'-1--

" ' 'I'-

' 1'-, 'I'- ' 1,

1' " 1' ' 1'

' ' 1'

' '\

' ' ' 1'

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•

•

NAC-STC SAR


June 2018

Revision 18B

Figure 3.4-34 Temperature Results for the Personnel Barrier

Pl P2

Cl

Cask Surface

C2 _/

C3

Boundary Conditions (°F)

Location Cl C2 C3 Pl

Temperature 230 244 258 140

3.4-79

P3

P4

PS

Personnel Barrier

Calculated Temperature (°F)

P2 P3 P4

105 126 127

PS

JOO

NAC-STC SAR


June 2018

Revision l 8B

Table 3.4-1 Maximum Component Temperatures-Normal Transport Conditions, Maximum

Decay Heat and Maximum Ambient Temperature-Directly Loaded and

Canistered Configurations

Conditions: 100°F Ambient Temperature, Full Insolation, Decay Heat Load: 22.1 kW for Uncanistered Fuel; 12.5

kW for Yankee-MPC Canistered Fuel; 17 kW for Connecticut Yankee-MPC Canistered Fuel

Canistered Fuel

Directly Loaded Yankee-MPC CY-MPC

Cavity Gas Cavity Gas Cavity Gas

Component Air (°F) Helium (°F) Notes Helium (°F) Notes Helium (°F) Notes

Outer Lid 0-ring 178 176 (1) 176 (4) 157 (6)

Port Cover 0-rings 211 210 (1) 210 (4) 179 (7)

Inner Lid and Port Cover

Plate 0-rings 190 189 (1) 189 (4) 179 (7)

Cask Radial Outer Surface 241 243 (2) 243 (5) 258 (8)

Top Neutron Shield 181 175 (1) 175 (4) 168 (8)

Radial Neutron Shield 284 285 (2) 270 (5) 288 (8)

Lead Gamma Shield 314 315 (2) 281 (5) 300 (8)

Aluminum Disk Exterior 338 337 (2) --- --- 331 (8)

Aluminum Disk Interior 491 487 (2) 536 (5) 534 (8)

Steel Support Disk Exterior 356 344 (2) --- --- 324 (8)

Steel Support Disk Interior 498 495 (2) 539 (5) 536 (8)

Canister Shell --- --- --- 338 (5) 351 (8)

Canister Lid --- --- --- 209 (5) 220 (8)

Canister Bottom Plate --- --- --- 255 (5) 347 (8)

Maximum Fuel Rod Cladding 588 544 (3) 575 (5) 611 (8)

Notes: (1) Temperatures are determined from the analysis of the three-dimensional quarter symmetry model of the entire cask (directly loaded fuel).

(2)

(3)

(4)

(5) (6)

(7)

(8)

Temperatures are determined from the analysis of the three-dimensional 180-degree section model of the entire cask ( directly loaded fuel). Temperatures are determined from the analysis of the two-dimensional detailed model of the fuel assembly (directly loaded fuel). Component not explicitly modeled in the 3-D model for Yankee-MPC canistered fuel. Temperature results from the helium case of the directly loaded fuel used (conservative).

Temperatures are determined from the 3-D model for Yankee-MPC canistered fuel. Not explicitly modeled-maximum temperature of cask outer lid from 3-D model for CY-MPC presented. Not explicitly modeled-maximum temperature of the cask top forging/cask lids from 3-D model for CY-MPC presented. Temperatures are determined from the 3-D model for CY-MPC canistered fuel.

3.4-80

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NAC-STC SAR


June 2018

Revision 18B

Table 3.4-2 Maximum Component Temperatures - Normal Transport Conditions, Maximum

Decay Heat, Minimum Ambient Temperature - Directly Loaded and Canistered

Configurations

Conditions: -40°F Ambient Temperature, No Insolation, Decay Heat Load: 22.1 kW for Uncanistered Fuel; 12.5 kW

for Yankee-MPC Canistered Fuel; 17 kW for Connecticut Yankee-MPC Canistered Fuel

Directly Loaded Yankee-MPC CY-MPC

Components Air (°F) Notes Helium (°F) Notes Helium (°F) Notes

Outer Lid 0-ring 125 --- 125 (2) 86 (4)

Inner Lid and Inner Lid Port 125 --- 125 (2) 89 (5)

Cover Plate 0-rings

Port Cover 0-rings 129 --- 129 (2) 170 (5)

Cask Radial Outer Surface 144 --- 116 (3) 162 (6)

Top Neutron Shield 131 --- 131 (2) 87 (6)

Radial Neutron Shield 181 --- 142 (3) 162 (6)

Lead Gamma Shield 215 --- 154 (3) 175 (6)

Fuel Basket Exterior 256 (1) --- --- --- ---Maximum Basket Web 399 (1) 431 (3) 428 (6)

Canister Shell --- --- 215 (3) 232 (6)

Canister Lid --- --- 71 (3) 92 (6)

Canister Bottom Plate --- --- 121 (3) 229 (6)

Fuel Rod Cladding 488 (1) 473 (3) 512 (6)

Notes: (1) Basket and fuel rod cladding temperatures are defined by adding the gradient result between the lead gamma shield and point of interest obtained from the 3-D directly loaded fuel model with air in the cavity (Table 3 .4-1 ).

(2) Component not explicitly modeled in the 3-D model for Yankee-MPC canistered fuel. Temperature results from the air case of the directly loaded fuel used (conservative).

(3) Temperatures obtained from 3-D model for Yankee-MPC canistered fuel. (4) Not explicitly modeled-maximum temperature of cask outer lid from 3-D model for Connecticut

Yankee-MPC presented. (5) Not explicitly modeled-maximum temperature of cask top forging/cask lids from 3-D model for

Connecticut Yankee-MPC presented. (6) Temperatures obtained from 3-D model for Connecticut Yankee-MPC canistered fuel.

3 .4-81

NAC-STC SAR


June2018

Revision J SB


Maximum Decay Heat, Low Ambient, for Directly Loaded Fuel

Conditions: -20°F Ambient Temperature, 22.l kW Decay Heat and No Insolation

Component Temperature (°F)

Outer Lid 0-ring 161

Inner Lid and Port Cover Plate 0-rings 165

Port Cover 0-rings 165

Cask Radial Outer Surface 173

Top Neutron Shield 168

Radial Neutron Shield 211

Lead Gamma Shield 245

Fuel Basket Exterior1 286

Maximum Basket Web 1 429

Maximum Fuel Rod Cladding1 518

1 Basket and fuel rod cladding temperatures are defined by adding the gradient

result between the lead gamma shield and point of interest obtained from the

3-D finite element analysis with air in the cavity (Table 3.4-1).

3.4-82

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NAC-STC SAR


June 2018

Revision 18B

Table 3.4-4 NAC-STC Thermal Performance Summary for Normal Conditions of Transport

Temperature Range

Directly Loaded Canistered Fuel

Component Fuel Yankee-MPC CY-MPC Allowable Temperature

< 380°C-Uncanistered

Fuel Cladding1 309°c 302°c 322°c < 340°C-Yankee-MPC

< 341 °C-CY -MPC

Metallic 0-rings -40 to 190°F -40 to 190°F -40 to 218°F -40 to 500°F

Viton 0-rings -40 to l 90°F -40 to 190°F -40 to 218°F -40 to 400°F

PTFE O-rings2 -40 to 190°F -40 to l 90°F -40 to 218°F -40 to 650°F

Radial NS-4-FR Neutron

Shield -40 to 285°F -40 to 270°F -40 to 288°F -40 to 300°F

Lead Gamma Shield -40 to 3 l 5°F -40 to 281 °F -40 to 300°F -40 to 600°F

Aluminum Heat Transfer Disk -40 tO 491 Of -40 to 536°F -40 to 534°F -40 to 600°F

I. Allowable temperatures for uncanistered ( directly loaded) fuel and for Yankee fuel in the

Yankee-MPC are based on PNL-4835. The allowable temperature for Connecticut Yankee

fuel is based on the methodology of PNL-6364, to consider preferentially loaded

configurations of the CY-MPC system.

2. The safe operating range extends to 735°F. (Certified Test Report D9-3362-I, Applied

Technical Services, Inc., February 8, 1989.) An allowable temperature of 650°F is

conservatively applied .

3.4-83



Table 3.4-5

NAC-STC

Components

Bottom Plate

Bottom Forging

Transition Shell

Inner Shell

Outer Shell

Top Forging

Inner Lid

Outer Lid

Inner Lid Bolt

Outer Lid Bolt

Notes: (I)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

Maximum Cask Component Temperatures in Normal Conditions of Transport

Directly Loaded Fuel Canistered Fuel

Cavity Gas Yankee-MPC CY-MPC

Air (°F) Helium (°F) Notes Helium (°F) Notes Helium (°F) Notes

350 333 (I) 333 (3) 347 (5)

417 393 (I) 393 (3) 347 (6)

300 300 (I) 300 (3) 331 (7)

331 331 (2) 311 (4) 331 (5)

292 293 (2) 276 (4) 294 (5)

211 210 (I) 210 (3) 218 (5)

223 210 (I) 210 (3) 217 (5)

178 179 (I) 176 (3) 216 (5)

190 189 ( 1) 189 (3) 217 (8)

178 176 ( 1) 176 (3) 216 (8)

Temperatures are determined from the analysis of the three-dimensional quarter symmetry model of

the entire cask.

Temperatures are determined from the analysis of the three-dimensional 180-degree section model

of the entire cask.

Component not explicitly modeled in the 3-D model for Yankee-MPC canistered fuel.

Temperature results from the helium case of the directly loaded fuel used (conservatively).

Temperatures are determined from the 3-D model for Yankee-MPC canistered fuel.

Temperatures are determined from the 3-D model for CY-MPC canistered fuel.

Not explicitly modeled-taken as the maximum temperature of the bottom plate from the 3-D

model for CY-MPC canistered fuel.

Not explicitly modeled-taken as the maximum temperature of the inner shell from the 3-D model

for CY-MPC canistered fuel.

Not explicitly modeled-taken as the maximum temperature of the inner and outer lids from the

3-D model for CY-MPC canistered fuel.

3.4-84

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NAC-STC SAR


June 2018

Revision 18B

Table 3.4-6 Maximum Component Temperatures for Yankee-MPC Damaged Fuel

Contents Heat Transfer Disk Damaged Fuel Fuel Cladding

Four damaged fuel and 32 548 544 586

intact fuel assemblies (°F)

Allowable Temperatures 600 NIA 644

(Table 3 .4-4) (°F)

Table 3.4-7 Maximum Component Temperatures for CY-MPC Damaged Fuel

Maximum Temperature1 (°F)

Design Damaged Fuel Reconfigured Fuel Fuel Rod

Condition Can Assembly Cladding2

100°F Ambient 611 611 611

-40°F Ambient 512 512 512

Bounding temperatures are taken from the maximum fuel cladding temperature for the design basis fuel

(654 watts per assembly) .

2 The allowable fuel cladding temperature is 646°F (341 °C) (See Table 3 .4-4).

Table 3.4-8 Westinghouse 15 x 15 Fuel Assembly Characteristics

Parameter Units Value

Number of Fuel Rods -- 204

Fuel Rod Outer Diameter inch 0.422

Fuel Pellet Diameter inch 0.3659

Fuel Rod Clad Inner Diameter inch 0.3736

Fuel Rod Length inch 152.756

Active Fuel Length inch 144.0

Fuel Rod Free Volume inch3 1.30

End Fitting Volume inch3 132.8

Grid Spacer Volume inch3 220.3

Guide/Instrument Tube Volume inch3 71.4

Spring Mass gram/rod 20.0

Fuel Rod Fill Pressure at Manufacture, 20° C psig 500

Moles of Backfill Gas moles/rod 0.031

3.4-85

NAC-STC SAR


Table 3.4-9 Directly Loaded Fuel Basket Component Volumes

Parameter Units

Bottom Weldment inch3

Top Weldment inch3

Support Disks inch3

Spacers inch3

Split Spacers inch3

Threaded Tie Rods inch3

Heat Transfer Disks inch3

Fuel Tubes inch3

Total inch3

3.4-86

June 2018

Revision 18B

Value

2,330

3,565

26,412

1,325

1,896

1,754

20,980

19,185

77,448

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NAC-STC SAR

Docket No. 71-9235

March 2017

Revision 18

How to Specify 0-Rings Denotes

Metallic 0-Rlng

I • Tubing 00 ; I Wall Thlclmeu

{Thirty-Seconcts) ! (Thouundlhs)

I ' II BIIII-IIE • II Materials

1-AlloV 7'18 7 -Stainless 2-Staintea Sleel 304

Steel 32, 8-Stainless 3-Alum-inum St111111316 4-Copper 9-Slainlllss 5-AUC)J 600 Steel 347 6-Allot~750 X-AsSpecified

Exllmpl•: U2312-0362SSEA The &bDve •umple. U2312-0J62~SEA. tnalcaln a type 321 atalnlea!I &IHI Cl-Ring. ~· /2.38 mm) tuba size . . 012 C0.30 mm) wall tt,icl<neu, 3.825' (112.00 mm) 00. aell-energlze<i (ID) ond ,001-.002" (0.03/0.0S mm) ,11,..., c:oaling.

t

Metallic Q.Ring OD {lncnes) (Thauaandths)

'Tyµ. SE-Satl-energized

on ID PF-Prenure filled NP-Not telf

-rgized, not prassure filled

SO--Sttll-ene rgizitd on OD

SX-Sell-ene rgized uspec.

Coatings A-Sil.,., .001 I .002 C0.03/0.051 N--None 8-Sllver .002/ .00310.05/0.08) P-Lead' .001 /.002 (0.03/0.051 D-Tellon• .001 / .003 (0.03/0.08) Fl-Indium .001 / .002 C0.03/0.05) E-Teflon .003/ .004 10.oaro.10) T-Nlckel .0011.002 ro.03/0.05) L-Copper .001 I .002 (0,03/0.05) V-Gold .0005/.001 (0,02/0.03)

X-AaSpec:~d

Fluorocarbon Metallic C-Rings Fluon:>carbc,n Metarnc c-Rings (designated MCA) 11re designed for stllllic: sealing on machinery or equipment and Bre 11vailllble for internal pressure, external preuu~. or axial pressure ID/OD applications. Because C-Rir,gs are desrgned with 11n open side on the pressure side of tile installation. the seal is &elf-energizing. Fluorocarbon C..flinps are offered in round or irregular snapes in a broad range 01 mes trom .126'" (3,2 mm) OD x .Da2" [0,81 mm) fr,11'! ~ight to ov-er 300"' (7620 mm) OD x 2-- (50.80 mm) free neigtlt. They are avail.ltlie in a wide variety of metal alloys and metallic: or Teflon coatinQS. SeaJing application temperature range is trom cryogenic to 3,000" F. 1iesoe C.); p,essur& tolerar,ces are from 10- torr to 100,000 psi (6.804 atm). Where customer requirnments 81'8 large. the C-Ring provides the lowest unit price of any high per1orrnance seal on the market.

•T""""d OuPDM~ ~T-.

@Helimflex Components Division Telephone (803) 783-1880 P.O. Box SB89 FAX 1803) 783-4279 Columbia. South Carolina 29290

4.5-13

NAC-STC SAR

Docket No. 71-9235, Revision l 7C

4.5.2 Blended Polytetrafluoroethylene (PTFE) 0-rings

June2018

Revision 18B

This section contains applicable technical data from a typical manufacturer of blended

polytetrafluoroethylene (PTFE) 0-rings. The PTFE 0-rings are manufactured from virgin

(unreprocessed) polytetrafluoroethylene base material filled with plastic. One product that

satisfies the design requirements is the Fluoroloy K 0-ring manufactured by the Furon Company,

which has an operating temperature range of -450°F to +650°F. NAC has completed

supplemental 0-ring testing and has determined that the operating range of the PTFE 0-rings can

be extended to 735 °F. A description of tests performed and the results are contained in Ce1tified

Test Report D9-3362-1, Applied Technical Services, Inc., February 8, 1989. Another product

that satisfies the design requirements is Parker Compound VM835-75. The compound's

recommended operating temperature range is -40°F to 400 °F.

4.5-14

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NAC-STC SAR

Docket No. 71-9235

March 2017

Revision 18

biasing, the limiting shift scenario corresponds exactly to the position of the fuel assembly in the

cavity, i.e., top dose rates are maximized when the fuel assembly is shifted up and bottom dose

rates are maximized when the fuel assembly is as far down in the cavity as possible. For radial

biasing, the two different shift scenarios are limiting for different transport conditions. The

maximum fuel assembly axial shift is limiting for normal conditions because upper plenum and

upper end fitting hardware move adjacent to the location in the radial shield where the radial lead

shield ends. The limiting shift is downward for accident conditions due to the bottom axial lead

slump, which is adjacent to the lower end fitting hardware source.

The first step in determining limiting PWR dose rates for the directly loaded cask is the

generation of dose rate response functions for generation of minimum cool time tables. For each

array size, at each of 4 burnup, 15 enrichment, and 18 cool time combinations, dose rate profiles

are calculated for both normal and accident transport conditions. Using these dose rate profiles,

the maximum radial dose rates at 2 meters from the railcar are tabulated for normal conditions.

Minimum cool times are calculated to ensure that a decay heat limit of 850 W/assembly is not

exceeded and that the dose rate at 2 meters from the railcar does not exceed 9.5 mrem/hr. The

9 .5 mrem/hr analysis limit was chosen to provide margin against the 10 mrem/hr regulatory limit.

Cool times needed to reach these limits are calculated using linear interpolation on the entire

array of maximum dose rates. The linear interpolation is valid because of the exponential

decrease in source term and, thus, dose rate as a function of time. The interpolated cool time is

rounded up to the next integer year. A sample minimum cool time generation for the 14xl4

reference assembly at 40,000 MWd/MTU is shown in Table 5.4-3. Repeating this analysis for all

fuel types and burnups results in the complete loading table shown in Table 5.4-5. Based on the

loading table, maximum radial dose rates for each fuel type are shown in Table 5.4-4.

The minimum cool times are used to calculate maximum accident condition dose rates at 1 meter

from the cask. The I 000 mrem/hr limit is not exceeded at any of the calculated minimum cool

times.

Based on the radial dose rate results for normal and accident conditions and their application to

the minimum cool time table, the 14x 14 reference assembly provides maximum dose rates. Thus,

top axial and bottom axial response functions have been executed for this assembly only. This

ensures that the maximum axial dose rates for the directly loaded system are captured, and that

variations in burnup, enrichment and minimum cool time are thoroughly examined .

5.4-3

NAC-STC SAR


June 2018

Revision 18B

A summary of the limiting source terms for each transport condition and detector biasing is given

below. All I imiting source terms are taken from the J 4x 14 reference fuel assembly.

Normal Conditions Accident Conditions

Detector Burnup Enrichment Cool Time Burnup Enrichment Cool Time

Biasing [MWd/MTU] [wt% 235U] [Years] [MWd/MTU J [wt% 235UJ [Years]

Radial 40,000 2.3 10 45,000 2.3 14

Top Axial 30,000 2.3 6 45,000 2.3 14

Bottom Axial 40,000 2.3 10 45,000 2.3 14

Three-Dimensional Dose Rates for Directly Loaded Fuel

Further detail on the three-dimensional dose rates are presented in Figures 5 .4-1 through 5 .4-6 for

the limiting 14xl4 reference assembly. Maximum dose rates are tabulated in Tables 5.4-6 and

5.4-7.

The maximum normal conditions surface dose rate is 366 mrem/hr at an axial elevation between

the radial neutron shield and the upper impact limiter. At l meter from the surface of the neutron

shield shell, the maximum dose rate is 20.3 mrem/hr. This dose rate defines the transport index.

The maximum normal conditions dose rate at 2 meters from the cask railcar is 9.5 mrem/hr and

occurs at an axial elevation adjacent to the upper plenum and upper end-fitting elevations. The

maximum accident conditions dose rate at 1 meter from the cask is 665 mrem/hr and occurs at

the cask midplane. The top and bottom axial dose rates are small when compared to the radial

dose rate for the same transport conditions.

Dose rate variations from heat fins in the neutron shield are examined explicitly using azimuthal

detectors that span the entire length of the neutron shield. As shown in Figure 5.4-4, peaks in the

neutron dose rate correspond to dips in the gamma dose rate, and vice versa. Thus, the neutron

dose rate increase resulting from the ducting is offset by the reduction of the gamma dose rate

resulting from the additional shielding provided by the fins. The use of thicker heat fins, which

have a pre-bonded thickness of 8mm to 10mm for 304 stainless steel and 6mm to 8mm for

copper plates, is acceptable as the increased heat fin thickness results in neutron doses that are

within the statistical uncertainty of the existing shielding analysis (i.e.,< 1 o}

Detector descriptions for dose rates on the side of the STC are given in Tables 5.4-8 and 5.4-9 for

normal and accident conditions, respectively. Note that an axial height of 0.0 cm corresponds to

the bottom of the STC cavity.

5.4-4

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•



Table of Contents

8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ................................. 8-1

8.1 Fabrication Requirements and Acceptance Tests ......................................................... 8.1-1

8.1.1 Weld Procedures, Examination, and Acceptance ............................................. 8.1-1

8 .1.2 Structural and Pressure Tests ............................................................................ 8.1-4

8.1.3 Leakage Tests .................................................................................................... 8.1-6

8.1.4 Co1nponent Tests .............................................................................................. 8.1-9

8.1.5 Tests for Shielding Integrity ........................................................................... 8.1-12

8.1.6 Thermal Test ................................................................................................... 8.1-15

8.1.7 Neutron Absorber Tests for NAC-STC Directly-Loaded Fuel Basket and

for Yankee-MPC and CY-MPC Canistered Fuel Baskets .............................. 8.1-18

8.1.8 Neutron Absorber Tests for MPC-LACBWR Canistered Fuel Basket .......... 8.1-20

8.1.9 Transportable Storage Canister ....................................................................... 8.1-22

8 .1.10 HL W Overpack and Basket ............................................................................ 8 .1-24

8.1.11 Alternative Neutron Absorber/Poison Tests for NAC-STC Directly

Loaded Basket. ................................................................................................ 8.1-27

8.2 Maintenance Program ................................................................................................... 8.2-1

8.2.1 Structural and Pressure Tests of the Casie. ....................................................... 8.2-1

8.2.2 Leak Tests ......................................................................................................... 8.2-2

8.2.3 Subsystems Maintenance .................................................................................. 8.2-3

8.2.4 Valves, Rupture Disks and Gaskets on the Containment Vessel.. .................... 8.2-4

8.2.5 Shielding ........................................................................................................... 8.2-4

8.2.6 Post-Fabrication Thermal Test. ......................................................................... 8.2-4

8.2.7 Miscellaneous ................................................................................................... 8.2-5

8.2.8 Maintenance Program Schedule ....................................................................... 8.2-6

8.3 Quick-Disconnect Valves ............................................................................................. 8.3-1

8.4 Cask Body Fabrication .................................................................................................. 8.4-1

8.4.1 General Fabrication Procedures ........................................................................ 8.4-1

8.4.2 Description of Lead Pour Procedures (Standard Method) ................................ 8.4-5

8.4.3 Description of Lead Pour Procedures (Alternate Method) ............................... 8.4-8

8-i

NAC-STC SAR


List of Figures

June2018

Revision 18B

Figure 8.1-1 Thermal Test Arrangement ............................................................................. 8.1-26

Figure 8.4-1 Typical Arrangement of Lead Pour Equipment (Standard Method) .............. 8.4-12

Figure 8.4-2 Typical Arrangement of Lead Pour Equipment (Alternate Method) .............. 8.4-13

List of Tables

Table 8.1-1 Neutron Absorber Material Minimum 10B Loading (NAC-STC Directly

Loaded Basket) ............................................................................................... 8.1-38

Table 8.1-2 Mechanical Properties of Neutron Absorber (NAC-STC Directly Loaded

Basket) ............................................................................................................ 8.1-38

Table 8.2-1 Maintenance Program Schedule ....................................................................... 8.2-7

8-ii

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NAC-STC SAR

Docket No. 71-9235

March 2017

Revision 18

will be applied in a vertical direction and equally distributed between the two rotation trunnion

recesses by the use of hydraulic rams combined with a load-spreading beam.

Following completion of the rotation trunnion recesses load test, all accessible trunnion recess

welds and load bearing surfaces shall be visually inspected for permanent deformation, galling or

cracking. Inspections utilizing liquid penetrant examination shall be performed in accordance

with the ASME Code, Section V, Article 6. Liquid penetrant acceptance standards shall be as

indicated in paragraph NF-5350 of the ASME Code, Section III, Division I.

Any evidence of permanent deformation, cracking, galling of the load bearing surfaces or

unacceptable dye penetrant results shall be cause for rejection of the rotation trunnion recesses or

related welds.

8.1.2.3 Hydrostatic Testing

A hydrostatic test shall be performed on the NAC-STC cask containment boundary, prior to final

acceptance of the cask, in accordance with the ASME Code, Section III, Division I, Article

NB-6200. The hydrostatic test pressure shall be at least 76 psig, which is 150 percent of the

Maximum Normal Operating Pressure. This test shall be performed in accordance with approved

written procedures. All pressure retaining components, appurtenances, and completed systems

shall be pressure tested.

The vent port will be used for the test connection. Only the vent port quick-disconnect will be

installed during the testing. The hydrostatic test will be performed with the inner lid and the

drain port coverplate installed and torqued.

The hydrostatic test system components, although not part of the cask containment boundary,

will be visually inspected prior to the start of the hydrostatic test. Leakage from the valves or

connections will be corrected prior to the start of the hydrostatic test.

The test pressure gauge installed on the cask will have an upper limit of approximately twice that

of the test pressure. The hydrostatic test pressure shall be maintained for a minimum of 30

minutes, during which time a visual inspection is made to detect any evidence of a leak. Any

evidence of a leak during the minimum hold period will be cause for rejection .

8.1-5

NAC-STC SAR


June 2018

Revision 18B

After completion of the hydrostatic test, the cask containment boundary will be dried and

prepared for visual and/or dye penetrant inspections as appropriate. The components of the cask

containment boundary shall be visually inspected. All accessible welds within the cavity shall be

liquid penetrant inspected. Any evidence of cracking or permanent deformation is cause for

rejection of the affected component.

8.1.2.4 Pneumatic Bubble Testing of the Neutron Shield Tank

A pneumatic bubble test of the neutron shield tank will be performed in accordance with Section

V, A11icle 10, Appendix I, of the ASME Code following final closure welding of the bottom

closure plates. The pneumatic test pressure shall be 12.5 + 1.5/-0 psig, which is 125 percent of

the relief valve set pressure. The test shall be performed in accordance with approved written

procedures.

During the test, the two relief valves on the neutron shield tank will be removed. One of the relief

valves threaded connections will be used for connection of the air pressure line and test pressure

gauge. The other relief valve connection will be plugged with a threaded plug.

•

Following introduction of pressurized air into the neutron shield, a 15-minute minimum soak •

time will be required. Following completion of the soak time, approved soap bubble solution will

be applied to all fin to shell, shell to end plate, and end plate to outer shell welds. The acceptance

criteria for the bubble test will be no air leak from any tested weld as indicated by continuous

bubbling of the solution. If an air leak is indicated, the weld shall be repaired in accordance with

approved weld repair procedures and the pneumatic bubble test shall be repeated until no

unacceptable air leak is observed.

8.1.3 Leakage Tests

Fabrication leakage rate testing is performed on both the NAC-STC transpo11 cask containment

boundary weldment (without the inner lid and inner lid vent and drain port coverplates installed)

during fabrication prior to lead pouring (without the inner lid and inner lid vent and drain port

coverplates installed) upon completion of cask body fabrication (i.e., following lead pouring and

final cask assembly) to demonstrate that the containment boundary weldment, as fabricated, will

provide an appropriate containment capability. The inner lid and with the inner lid vent and

drain port coverplates installed will be leakage rate tested as part of the final fabrication leakage

rate testing per SAR Section 8.1.3.2.

8.1-6 •

• NAC-STC SAR


June 2018

Revision 18B

The leakage rate testing of the NAC-STC containment boundary and closures will be performed

in accordance with the requirements and standards contained in ANSI N 14.5-1997 and the

ASME Code, Section V, Article 10. Helium leakage test procedures shall be developed and

approved by personnel qualified in accordance with the requirements of SNT-TC-1A as a NDT

Level III (Leak Testing) examiner. The leakage rate tests shall confirm that the leakage rate

meets the containment criteria established in Chapter 4 and Table 7.4- 1 ( e.g., leaktight for the

NAC-STC containment boundary weldment; leaktight for Containment Condition A containment

boundary closures with metallic seals for spent fuel transport following long-term storage;

leaktight for Containment Condition BI closures with metallic seals for directly loaded PWR

spent fuel assemblies with burnups of :S 45 GWd/MTU, and for canistered spent fuel and GTCC

waste, and HL W overpacks; leaktight for Containment Condition B3 closures with Viton 0-ring

or metallic containment seals for HBU spent fuel assemblies; or Containment Condition B2 with

Vi ton 0-ring or metallic containment seals with cumulative leakage rate :::; 9 .3 x I 0-5 cm3 /sec

(helium) for containment boundary components with Viton 0-ring seals and leaktight for

metallic containment seals for standard, directly loaded PWR spent fuel assemblies). Leak tests

shall be performed by personnel qualified for helium leakage testing in accordance with the

requirements of ANSI/ASNT CP-189-2006, "Standard for Qualification and Certification of

• Nondestructive Testing Personnel."

8.1.3.1 Containment Boundary Weldment Fabrication Leakage Rate Test

Following the satisfactory completion of hydrostatic pressure testing of the NAC-STC

containment boundary weldment per Section 8.1.2.3, the containment vessel cavity is drained and

cleaned. Per Paragraph 7.3 of ANSI N 14.5-1997, a helium fabrication leakage rate test of the

containment boundary weldment will be performed in accordance with the requirements of

Section V, Article IO of the ASME Code. The containment boundary weldment shall be leakage

tested to demonstrate a leak rate of less than, or equal to, 2x I 0-7 cm3/sec (helium) with a

minimum test sensitivity of Ix 10-7 cm3/sec (helium) to verify the containment boundary

weldment, including containment welds and base materials, is leaktight as defined in ANSI

Nl4.5-1997.

If a leak exceeding the leakage rate acceptance criteria is detected, the affected weld or area of

base metal shall be rejected. Rejected welds or areas of base metal shall be repaired in

accordance with the requirements of Article NB-4450 of the ASME Code. The repaired weld or

base metal area shall be reexamined using the same procedure and acceptance criteria as

specified for the original weld examination. The helium fabrication leakage rate test of the

containment boundary weldment shall then be re-performed in accordance with the original test

• requirements and acceptance criteria prior to final acceptance.

8.1-7

NAC-STC SAR

Docket No. 71-9235

8.1.3.2 Final Fabrication Leakage Rate Testing

March 2017

Revision 18

Upon completion of cask body fabrication, a helium fabrication leakage rate test is performed on

the removable containment boundary closure components and their respective metallic seals or

Viton 0-ring seals (i.e., cask inner lid, lid bolts, inner lid vent and drain port coverplates and

bolts) in accordance with Paragraph 7.3 of ANSI Nl4.5-1997.

• Final containment fabrication leakage rate testing is performed with the cask assembled in

accordance with the cask assembly drawing, except that the vent or drain quick-disconnect is not

installed (Note: The test is repeated to ensure that both the vent and drain port coverplates are

individually leakage tested to the applicable criteria). This ensures that when the cask cavity is

backfilled with helium, helium is present on the containment side of the vent or drain port

coverplate containment 0-ring. Leakage rate tests are performed on the cask lid and the lid pmi

coverplate and their respective 0-ring seals. The test is performed using a helium mass

spectrometer leak detection system by establishing a vacuum of :S 0.1 torr in the seal or 0-ring

annulus of the cask lid and, separately, of the vent and drain port coverplate using the applicable

test port. The cask containment boundary is backfilled with a known concentration of high purity

helium gas. The acceptance criteria for the containment fabrication leakage rate testing of the

removable closures are provided in Section 8.1.3.2.1 and 8.1.3.2.2, depending on the seal •

material and type (e.g., metallic seals or Viton 0-rings). A leakage rate that exceeds the

allowable leakage rate is cause for rejection of the component and seal being tested. Seal

replacement or other corrective actions shall be taken to repair any detected leaks. The

component and replaced seal shall then be retested and re-inspected in accordance with the

original test requirements and acceptance criteria prior to final acceptance. After successful

completion of the leakage tests, quick-disconnects are installed in the inner lid vent and drain

port openings and torqued.

8.1.3.2.1 Metallic Seal Testing Acceptance Criteria

The fabrication leakage rate testing of the containment boundary closures using metallic seals

consists of a series of leak tests. The acceptance criteria for each metallic seal is a detected

leakage rate of :S 2 x 10-7 cm3/sec (helium) at a test system sensitivity of< 1 x 10-7 cm3/sec

(helium) or better.

8.1.3.2.2 Viton 0-Ring Testing Acceptance Criteria

The fabrication leakage rate testing of the containment boundary closures using Viton 0-rings

consists of a series of leak tests. The acceptance criteria for the Viton 0-ring seals is a

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NAC-STC SAR

Docket No. 71-9235

March 2017

Revision 18

cumulative (sum of the three individual leakage tests results) detected leakage rate of :'.S 9.3 x 10-5

cm3/sec (helium) at a test system sensitivity of < 4.7 x 10-5 cm3/sec (helium) or better for

standard directly loaded PWR spent fuel assemblies, or leak tight leakage rate of :'.S 2.0 x 10-7

cm3/sec (helium) at a test system sensitivity of< 1.0 x 10-7 cm3/sec (helium) or better for directly

loaded HBU PWR spent fuel assemblies.

8.1.4 Component Tests

Tests performed on individual components are designed to ensure that the component meets the

design requirements for correct and proper operation of the cask system.

Acceptance criteria are established based on the functions and design requirements of the

component being tested.

8.1.4.1 Valves

There are no valves that are part of the NAC-STC containment boundary for transport. Quick

disconnects are installed in the vent, drain and interseal test port openings in the inner lid to

provide access to the cavity, and in the interlid port to provide access to the interlid region.

These fittings serve as valves when the mating parts are connected, and are used to connect

ancillary equipment to the cask cavity for filling, draining, drying, backfilling, gas sampling, and

leak testing operations. Upon removal of the external fitting, the valve in the quick-disconnect

closes automatically. The design and selection of the quick-disconnects is based on similar

equipment and procedures used with other NRC-approved storage and transport casks. For

transport, the quick-disconnects are sealed inside the transport containment boundary using a

bolted coverplate fitted with two 0-ring seals.

There are no rupture disks on the NAC-STC.

Two self-actuating pressure relief valves are installed on the external shell of the neutron shield

to provide for venting of vapor from the shielding material during transport thermal accident

conditions. These valves have stainless steel bodies and an operating pressure range of zero to

200 psig with an adjustable cracking pressure within this range. The cracking pressure is set at

10 psig. These relief valves do not provide a safety function, but have been designed to

minimize recovery effo1is in the unlikely event of a neutron shield overpressure condition .

8.1-9

NAC-STC SAR


8.1.4.2 Gaskets

June 2018

Revision 18B

As described in Section 8.1.3, the containment boundary of the NAC-STC may use either

metallic 0-rings or non-metallic Viton 0-rings. The two 0-ring types require different 0-ring

groove designs and, therefore, may not be used interchangeably and must be used with the inner

lid, vent and drain port coverplates and outer lid having the appropriate 0-ring groove machined

in the component. Metallic 0-rings are required to be used for direct loading of the NAC-STC

with fuel for extended storage and for loading of a transportable storage canister (for transport).

For direct loading of fuel for immediate transpo1i, either metallic or non-metallic 0-rings may be

used.

The outer lid, inner lid, drain port coverplate, vent port coverplate, interlid port cover, pressure

p01i cover, and interseal test plug gaskets are 0-rings. For transport after an extended period of

storage, the containment boundary is formed by the outer metallic 0-ring of the inner lid, the

outer metallic 0-rings on the vent and drain port coverplates, and the interseal test plug metallic

0-rings for the inner lid, the vent port coverplate and the drain port coverplate. The inner

metallic 0-rings of the inner lid, vent port coverplate and drain port coverplate, the metallic

0-ring of the outer lid, and the Viton 0-rings of the interlid and pressure po1i covers provide a

secondary closure to the cask contents. For immediate transport, the containment boundary is

formed by the inner 0-rings of the inner lid and vent and drain port coverplates. A second

boundary is formed by the 0-rings of the outer lid and interseal and pressure port covers.

The 0-ring replacement schedule depends upon the 0-ring material. The metallic 0-ring(s) of

any component shall be replaced prior to reinstallation of the component during loading

operations. Metallic seal replacement is not required prior to the transport of an empty NAC

STC. Those Viton 0-rings that provide the Containment Boundary seal shall be replaced

annually during cask transport operations, or prior to transport if they have been installed longer

than one year. Secondary Boundary (i.e., Non-Containment Boundary) Viton 0-rings shall be

replaced at least once every two years during cask transport operations, or prior to transport if

they have been installed longer than two years.

The containment boundary 0-rings shall be tested and maintained in accordance with the

Maintenance Program Schedule of Table 8.2-1 and the leak test criteria of Section 8.2.2.

8.1-10

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NAC-STC SAR


8.1.4.3 Miscellaneous

June 2018

Revision 18B

The removable transport impact limiters consist of redwood and balsa wood. License drawings

and the supporting analyses specify the crush strengths of the redwood and balsa wood to be

6240 psi± 620 psi and 1550 psi± 150 psi respectively. For manufacturing purposes, verification

of the impact limiter material is accomplished by verifying the densities of the wood. Three

samples from each redwood board are to be tested for density, and the average density of the

samples shall be 23.5 ± 3.5 pounds/cubic foot. Each 15-degree and 30-degree pie shaped section

of the impact limiter shall have a density of 22.3 ± 1.2 pounds/cubic foot in accordance with the

License Drawings. The moisture content for any single redwood board must be greater than 5

percent, but less than 15 percent. The average moisture content for a lot of redwood used in

impact limiter construction must not be greater than 12 percent.

Following final closure welding of the transport impact limiter stainless steel shell, a leak test of

the shell welds shall be performed to verify weld integrity. The following are acceptable test

methods, which may be selected from to verify weld integrity:

1. A test may be performed by evacuating the impact limiter to 75 mbar and performing a

30-minute test to determine if there is any increase in the impact limiter pressure. Any

detected leak shall not exceed 1 x 10-2 cm3/sec. If a leak exceeding this value is detected,

the cause of the leak shall be determined, and the weld repaired and retested.

2. A positive pressure leak test may be performed on each impact limiter to ensure the leak

tightness of the impact limiter shell welds. Remove the test plug and install the necessary

piping to convey oil-free air or gas to the inside of the impact limiter shell. Apply an air

or gas pressure to the inside of the impact limiter shell to initiate the test. Allow the

system to stabilize for at least 15 minutes. Spray all the outside welds with foaming

bubble solution. Examine the limiter welds for indications of continuous bubble

formation. All leaks detected shall be repaired and the leak test re-performed until there

are no leak indications. Upon the completion of leak testing, the test plug shall be

reinstalled.

3. After final closure welding of the transport impact limiter stainless steel shells, a PT

examination may be performed on all shell welds to verify weld integrity. Liquid

penetrant examined per ASME B&PV Section V, Article 6. Acceptance per Section Ill,

Article NF-5350 .

8.1-11

NAC-STC SAR

Docket No. 71-9235, Revisions 17 A and l 7B

8.1.5

8.1.5.1

Tests for Shielding Integrity

Gamma Shield Test

June 2018

Revision 18B

The gamma scan test shall be conducted by continuous scanning or probing over 100 percent of

all accessible cask body surfaces, which directly shield regions where lead was poured, using a

detector and a 6°Co source. Accessible cask surfaces are not only those surfaces that are physically

accessible but also cask surfaces where accurate detector readings can occur. The source strength

shall be of an intensity sufficient to produce a count rate that equals or exceeds three times the

background count rate on the external surfaces of the cask. The count rate shall be maintained

for greater than one minute prior to the start of scanning. The detector scan path spacing ( cask

body exterior surface) will be sufficiently small such that there will be scanning overlap based on

the size detector used and the scanning speed will be 4.5 feet per minute or less. The source scan

path spacing ( cask interior surface) will be on a sized grid pattern that is sufficiently small such

that scanning overlap will occur based on the size detector used.

A gamma scan test is not required for the cask inner closure lid, cask outer closure lid, cask inner

bottom forging, cask outer bottom forging, or cask outer bottom plate. These components shall be

ultrasonic tested in order to demonstrate their soundness as gamma shielding. Ultrasonic testing

shall be performed per ASME B&PV NB-2542.1 using the acceptance standards of Section NB-

2542.2 for forgings and ASME B&PV NB-2532.1 using the acceptance standards of NB-

2532.1 (b) for plates.

The acceptance criteria for the cask body shield test shall be that the shield effectiveness of the

cask body is equal to or greater than the shield effectiveness of a lead and steel mock-up. The

steel thickness of the mockup shall be equivalent to the minimum steel thickness specified on the

License Drawings and the lead thickness shall be equivalent to the minimum lead thickness

specified in the License Drawings Jess 3 percent. The shielding mock-up will be produced using

the same fabrication techniques as those approved for the cask.

Measured count rates that exceed those established by the test mock-up shall cause the

component to be rejected. The rejected areas/components shall be evaluated to determine the

corrective action to be taken. Any repaired areas shall be retested prior to acceptance.

8.1-12

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NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

An additional gamma shield effectiveness test shall be performed on each cask following first

fuel loading. The neutron and gamma shield effectiveness test procedures and acceptance criteria

are described in Section 8.1.5.4.

8.1.5.2 Neutron Shielding Test

The neutron shielding of the NAC-STC is provided by a solid layer ofNS-4-FR, which is a hard

polymer material. A 5.5-inch layer of NS-4-FR is located in the annulus formed by the outer

shell and the 0.236-inch (6 mm) thick neutron shield shell. The neutron shield is divided in

sections by the copper/stainless steel fins. A 2-inch thick layer of NS-4-FR is also installed in

the cask inner lid and in the cask bottom.

The installation of NS-4-FR material in the fabrication of the cask is a special process and, as

such, procedures will be prepared and qualified to ensure that the mix ratios, mixing method,

degassing, pouring, and curing of the material is properly performed. The NS-4-FR raw material

is provided in the form of a 3-part mixing kit. The material content of the raw material is tested

and certified at the time of kit preparation. The neutron shielding material is installed into the

annulus between the outer shell and the neutron shield shell by pouring it with the cask in an

inverted vertical position. Prior to installation, samples from each mix of the actual material

being poured into the annulus are wet density tested to ensure that the material is properly mixed.

Mixes that do not meet the wet density acceptance criteria are rejected. Procedures used for

installation of the material are validated prior to use by destructive examination of a full scale

mock-up of the neutron shield cavity. Qualification of the installation procedure verifies material

homogeneous properties and minimizes the potential deleterious voids.

8.1.5.3 Neutron Shielding Material Testing

The neutron shield prope11ies of NS-4-FR are provided in Chapters I and 3. Each lot (mixed

batch) of neutron shield material shall be tested to verify that the hydrogen concentration, boron

concentration, and neutron shield density meet the requirements specified in Chapters 1 and 3

and the License Drawings. Testing shall be performed by qualified laboratories in accordance

with written and approved procedures. Hydrogen concentration, boron concentration, and

density data for each Jot of neutron shield material shall become part of the quality record

documentation package.

Dimensional inspection of the cavities containing the neutron shielding material shall ensure that

• the required thickness specified in the License Drawings is incorporated into the cask.

8.1-13

NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

The installation of the neutron shielding material shall be performed in accordance with written,

approved, and qualified procedures. The procedures shall ensure that mix ratios and mixing

methods are controlled in order to achieve proper material composition, boron concentration and

distribution, and that pours are controlled in order to prevent gaps or unacceptable voids from

occurring in the material. Procedures shall be qualified by the use of mock-ups to ensure that the

NS-4-FR installation does not result in the creation of unacceptable voids. Wet density data for

each mix of installed neutron shield material shall be maintained as part of the quality record

documentation package.

8.1.5.4 Neutron and Gamma Shield Effectiveness Tests

Following first fuel loading, a neutron and gamma shield effectiveness test shall be performed for

each cask prior to transport. The test shall be performed with the cask loaded with fuel, drained,

vacuum dried and backfilled with helium. The purpose of the test is to document the

effectiveness of the neutron and gamma shielding materials. The test shall be performed in

accordance with detailed, approved written test procedures.

Calibrated neutron and gamma dose rate meters shall be used to measure the neutron and gamma

dose rate at contact with the outer shell of the neutron shield and at 2.3 meters from the surface

(equivalent to 2 meters from the sides of the railcar). Dose measurement points shall be

established on the external surface of the shell at 30° intervals and at five points along the height

of the shield (a total of 60 measuring points). In addition, neutron and gamma dose rate

measurements shall be made of the trunnion areas above the neutron shield, at four points below

the neutron shield, and at the edges and center of the cask top ( outer I id) and cask bottom

surfaces. Dose rates at the top and bottom of the cask shall be measured with the transport

impact limiters installed. The dose rates measured at contact and at 2.3 meters shall be recorded

on the test data sheet, along with the total power of the loaded fuel assemblies; date, time and

location of test; identification and calibration of instrumentation; and identification of test

engineer and operators.

To allow an evaluation of the measured dose rates to be completed, the burn up and cool time for

the actual fuel assemblies loaded into the cask will be determined and recorded. From this fuel

history data, the total actual neutron and gamma source terms will be estimated using ORIGEN

or similar calculations.

8.1-14

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NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

If the measured dose rates exceed the applicable regulatory limits, the licensee shall notify the

NRC. Appropriate corrective measures will be taken, including fuel unloading and correction of

the shielding deficiency. Following corrective actions, the test will be re-performed to the

original acceptance criteria prior to final acceptance.

8.1.6 Thermal Test

Prior to acceptance at the factory, a thermal test shall be performed on each fabricated packaging

to confirm and verify that the fabricated and assembled cask possesses the heat rejection

capabilities predicted by the thermal analyses. The thermal test shall be performed in accordance

with approved written procedures.

8.1.6.1 Thermal Test Set-up

The thermal test set-up 1s shown in Figure 8.1-1 (a). As depicted, the thermal test shall be

performed with the cask positioned horizontally on a test frame. The transport impact limiter or

equivalent insulating material shall be installed on each end of the cask to simulate the transport

configuration. The cask will be located in a covered building in a still environment. The cask

shall be assembled with the basket installed. A thermal test lid with connections for

thermocouple leads and electric heater power cables shall be installed in place of the inner lid.

The outer lid will not be installed for the test. The thermal test lid will be provided with an 0-

ring seal capable of containing the containment cavity helium atmosphere.

Electric heaters shall be installed in each fuel tube. The electric heaters will have an active length

of between 120 and 150 inches and be capable of generating a minimum of 22 kilowatts (kw).

The heaters will be supported in the basket so as to not be in contact with the wall of the fuel

tube. The power supplied to the heater will be recorded throughout the test duration.

Calibrated test thermocouples, with an accuracy of ±2°F, will be installed on the cask basket,

inner shell, and outer neutron shield shell surfaces. The location of the test thermocouples are

shown in Figure 8.1-1. The specific location of the thermocouples are as follows:

TC 1 - basket top steel weldment

TC2 - steel disk at cask basket midpoint

TC3 - aluminum disk at cask basket midpoint

TC4 - basket bottom steel weldment

8.1-15

NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

TC5; TC6; TC7; and TC8 - located at 90° intervals on the inner shell surface at cavity

midpoint

TC9 - top of inner shell surface at 30-40 inches from top of cavity

TC 1 0 - bottom of inner shell surface at 3 0 to 40 inches from base of the cavity

TCI 1; TC12; TC13; and TC14 - located at 90° intervals on the neutron shield shell

surface (at fin tip) at cask midpoint

TC15 - top of neutron shield shell surface (at fin tip) at 30-40 inches from top of neutron

shell

TC16-bottom of neutron shield shell surface (at fin tip) at 30-40 inches from bottom of

neutron shield shell

TC 1 7 - top of upper forging

TC 18 - outer shell surface at centerline of cask bottom fac;:e

TC 19 - inner fuel tube wall surface near the center of the cask basket

TC20 - ambient temperature of testing area

The output of the test thermocouples will be recorded throughout the test by a strip chart

recorder.

8.1.6.2 Test Procedure

With the cask assembled and instrumented as described above, the cask cavity is evacuated and

backfilled to 1.0 atmosphere absolute (14.6 psia) with helium. Power will be applied to the

heaters to simulate the cask contents. After initiation of power to the heaters, the temperatures of

all thermocouples and heater power levels will be monitored and recorded on data sheets at 60

minute intervals. Power will be maintained to the electrical heaters until the cask has reached

thermal equilibrium.

For the purpose of the test, thermal equilibrium is defined as being achieved when over two

consecutive hours:

~tTCJ3::::; 2°F/hr, and

~tTc3::::; 2°F/hr

Based upon the thermal heat-up evaluation, thermal equilibrium should be achieved m

approximately five days.

8.1-] 6

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NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

After verification of thermal equilibrium, final temperature measurements will be recorded for all

test thermocouples. The final power readings for the electric heaters will also be recorded. The

strip chart will be marked to indicate the time of the final cask measurements. The printout of the

strip chart recorder and the completed test data sheets will be incorporated into an approved final

thermal test report. The test will be determined to be acceptable if the acceptance criteria of

Section 8.1.6.3 are met.

If the acceptance criteria are not met, the cask will not be accepted until appropriate corrective

actions are completed. Upon completion of corrective actions, the cask shall be retested to the

original test requirements and acceptance criteria.

8.1.6.3 Acceptance Criteria

The purpose of the thermal test is to confirm the heat rejection capabilities of the as-built cask

are acceptable and correspond to the temperatures calculated by thermal analyses for the directly

loaded (uncanistered) configuration presented in Chapter 3.0 of this application .

Package heat dissipation acceptance testing assures: 1) maximum material temperatures do not

exceed material allowables; and that 2) measured temperature gradients are less than the thermal

gradients calculated in the package thermal analyses.

The thermal acceptance test is accepted when the following criteria are met:

I) When corrected for physical test boundary conditions and heat load, the following

. measured temperatures are not exceeded:

TC No. Location

TCl Top Basket Steel Weldment

TC3

TC2

TC4

TC5-TC8

TCl l-TC14

TC17

TC18

TC19

Aluminum Disk Center

Steel Support Disk Center

Basket Bottom Steel Weldment

Cask Inner Shell


Cask Top Forging

Cask Bottom

Tube Wall

8.1-17

Temperature °F

435

485

495

475

330

240

200

330

540

NAC-STC SAR

Docket No. 71-9235

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Revision 18B

2) The measured temperature gradient across the central steel disk from TC2 to the

average of TC5, TC6; TC7 and TC8 is less than 200°F;

3) The measured temperature gradient across the central aluminum disk from TC3 to

the average of TC5; TC6; TC7 and TC8 is less than l 90°F; and

4) The measured temperature gradient across the cask body as measured by

thermocouple pairs TC5-TC13; TC6-TC14; TC7-TC11; and TC8-TC12 are less

than 90°F.

8.1.7 Neutron Absorber Tests for NAC-STC Directly-Loaded Fuel Basket and for

Yankee-MPC and CY-MPC Canistered Fuel Baskets

Two alternate neutron poison materials, BORAL and Ta!Bor, have been qualified by NAC for

use in the NAC-STC directly loaded, the Yankee-MPC and the CY-MPC fuel baskets. For the

NAC-STC directly loaded basket BORAL acceptance testing is described in this section (Section

8.1.7) while a generic metal matrix composite (aluminum based) and borated aluminum

acceptance/qua! ification program is described in Section 8.1 .11. The generic program is

designed to demonstrate structural, thermal, and nuclear requirements are met without

•

specification of a particular manufacturer or material. Ta!Bor, an MMC material, is excluded •

from the generic acceptance/qualification program, and is included in Section 8.1.7.

BORAL is manufactured by Ceradyne Corporation, Chicoutimi (Quebec), Canada under a

Quality Assurance/Quality Control program in conformance with the requirements of 10 CFR

50, Appendix B. The manufacturing process consists of several steps: the first step is the mixing

of the aluminum and boron-carbide powders that form the core of the finished material, with the

amount of each powder a function of the desired 10B areal density. The methods used to control

the weight and blend of the powders are patented and proprietary processes of AAR Advanced

Structures (AAR) (subsequently Ceradyne). The mixture of powders is placed in an aluminum

box with walls approximately one inch thick. The top lid is welded in place. This "ingot" is

heated for several hours and then is hot-rolled to produce the sheet of design thickness. The

rolling process densifies and bonds the powder mixture. The aluminum box walls become the

cladding for the Al-B4C core.

Ta!Bor is manufactured by Talon Composites, Inc. (Ta!Bor was formerly called Boralyn, and

was produced by Alyn Corporation. Alyn Corporation went out of business and Talon

Composites acquired the major production equipment and the patent rights for Boralyn. Ta!Bor

is essentially identical to Boralyn.) Ta!Bor is manufactured and controlled using a Quality

Assurance program that is compliant with the applicable requirements of 10 CFR 50, Appendix,

8.1-18 •

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NAC-STC SAR

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B. TalBor is a metal matrix composite (MMC). The aluminum and B4C powders are mixed to

the specified JOB areal density and the powder mixture is vacuum sintered and hot pressed to

achieve a fully dense billet. The billet is extruded, then cut and rolled to the design thickness.

After manufacturing, test samples from each batch of neutron absorber (poison) sheets shall be

tested to verify the presence, proper distribution, and minimum weight percent of JOB. Neutron

transmission testing or augmented wet chemistry testing may be used. The tests shall be

performed in accordance with approved written procedures.

8.1.7.1 Neutron Absorber Material Sampling Plan

The neutron absorber sampling plan is selected to demonstrate a 95/95 (95% probability and 95%

confidence level) statistical confidence level in the neutron absorber sheet material compliance

with the specification. In addition to the specified sampling plan, each sheet of material is

visually and dimensionally inspected using at least 6 measurements (along the edges near each

corner and the longitudinal centerline) on each sheet. No rejected neutron absorber sheet is used.

The sampling plan is supp01ied by written and approved procedures .

The sampling plan requires that a coupon sample be taken from each sheet of the first set of 100

sheets of absorber material. Thereafter, coupon samples are taken from 20 randomly selected

sheets from each set of 100 sheets. This 1 in 5 sampling plan continues until there is a change in

lot or batch of constituent materials of the sheet (i.e., boron carbide powder, aluminum powder,

or aluminum extrusion), or a process change, at which time the sampling process is reinitiated as

previously described. The sheet samples are indelibly marked and recorded for identification.

This identification is used to document neutron absorber test results, which become part of the

quality record documentation package.

8.1.7.2 Wet Chemistry Test Performance

An approved facility with chemical analysis capability shall be selected to perform the wet

chemistry tests. The tests will ensure the presence of boron and enable the calculation of the JOB

areal density. Acceptability of the uniformity of boron distribution is based on the

manufacturer's material qualification tests.

The most common method of verifying the acceptabi 1 ity of neutron absorber material is the wet

chemistry method-a chemical analysis where the aluminum is separated from a sample with

known thickness and volume. The remaining boron-carbide material is weighed and the areal

8.1-19

NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

density of 10B is computed. A statistical conclusion about the BORAL or Ta!Bor sheet from

which the sample was taken and that batch of sheets may then be drawn based on the test results

and the established manufacturing processes previously noted.

BORAL and Talbor sheets are required to contain a minimum 0.020 g 10B/cm2 which is credited

at 75% effectiveness in Chapter 6.

8.1.7.3 Neutron Absorption Transmission Test Performance

An approved facility with a neutron source and neutron detection capability shall be selected to

perform the described tests, if the neutron absorption transmission test method is used. The tests

will assure that the neutron absorption capacity of the material tested is equal to, or higher than,

the given reference value and will verify the uniformity of boron distribution. The principle of

measurement of neutron absorption is that the presence of boron results in a reduction of neutron

flux between the thermalized neutron source and the neutron detector-depending on the

material thickness and boron content.

Typical test equipment will consist of thermal neutron source equipment, a neutron detector and

a counting instrument. The test equipment is calibrated using a known standard, whose 10B

content has been checked and verified by an independent method such as chemical analysis. This

calibration process shall be repeated daily (every 24 hours) while tests are being performed.


The neutron transmission test results shall be considered acceptable if the minimum 10B areal

density is determined to be equal to, or greater than, that specified on the fuel tube drawings.

Any specimen not meeting the acceptance criteria shall be rejected and all of the sheets from that

batch shall be similarly rejected unless coupons from each individual absorber plate are tested

and confirmed to meet or exceed the specified areal density.

8.1.8 Neutron Absorber Tests for MPC-LACBWR Canistered Fuel Basket

Neutron absorber material ( commercially available as BORAL ®), in the form of sheets consisting

of boron-carbide evenly dispersed within a matrix of aluminum and clad with aluminum, is used

in the NAC-MPC transportable storage canister fuel baskets. The manufacturing process consists

of several steps - the first being the mixing of the aluminum and boron-carbide powders that

form the core of the finished material, with the amount of each powder a function of the desired

8.1-20

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108 areal density. The methods used to control the weight and to blend the powders were

patented and proprietary processes of AAR and, subsequently, of Ceradyne Corporation of

Chicoutimi (Quebec), Canada.

After manufacturing, test samples from each batch of BORAL ® neutron absorber (poison) sheets

shall be tested using wet chemistry or neutron absorption techniques to verify the presence,

proper distribution, and minimum weight percent of 10B. The tests shall be performed in

accordance with approved written procedures.

8.1.8.1 Neutron Absorber Material Sampling Plan

The neutron absorber sampling plan is selected to demonstrate a 95/95 (95% probability and 95%

confidence level) statistical confidence level in the neutron absorber sheet material

compliance with the specification. In addition to the specified sampling plan, each sheet of

material is visually and dimensionally inspected using at least six measurements (along the edges

near each corner and the longitudinal centerline) on each sheet. No rejected neutron absorber

sheet is used. The sampling plan is supported by written and approved procedures.

The sampling plan requires that a coupon sample be taken from each sheet of the first set of 50

sheets of absorber material. Thereafter, coupon samples are taken from 10 randomly selected

sheets from each set of 50 sheets. This I in 5 sampling plan continues until there is a change in

lot or batch of constituent materials of the sheet (i.e., boron carbide powder, aluminum powder,

or aluminum extrusion), or a process change, at which time the sampling process is reinitiated as

previously described. The sheet samples are indelibly marked and recorded for identification.

This identification is used to document neutron absorber test results, which become part of the

quality record documentation package.

8.1.8.2 Wet Chemistry Test Performance

An approved facility with chemical analysis capability shall be selected to perform the wet

chemistry tests. The tests will ensure the presence of boron and enable the calculation of the 10B

areal density. Acceptability of the uniformity of boron distribution is based on the

manufacturer's material qua! ification tests.

The most common method of verifying the acceptability of neutron absorber material is the wet

chemistry method - a chemical analysis where the aluminum is separated from a sample with

known thickness and volume. The remaining boron-carbide material is weighed and the areal

• density of 108 is computed. A statistical conclusion about the BORAL ® sheet from which the

8.1-21

NAC-STC SAR

Docket No. 71-9235

June2018

Revision 18B

sample was taken and that batch of BORAL ® sheets may then be drawn based on the test results

and the established manufacturing processes previously noted.

8.1 .8.3 Neutron Absorption Test Performance

An approved facility with a neutron source and neutron detection capability shall be selected to

perform the described tests, if the neutron absorption test method is used. The tests will assure

that the neutron absorption capacity of the material tested is equal to, or higher than, the given

reference value and will verify the uniformity of boron distribution. The principle of

measurement of neutron absorption is that the presence of boron results in a reduction of neutron

flux between the thermalized neutron source and the neutron detector-depending on the

material thickness and boron content.

Typical test equipment will consist of thermal neutron source equipment, a neutron detector and

a counting instrument. The test equipment is calibrated using standards whose 10B content has

been checked and verified by an independent method such as chemical analysis. The highest

permissible counting rate is determined from the neutron counting rates of the reference sheet(s),

•

which should be ground to the minimum allowable plate thickness. This calibration process shall •

be repeated daily (every 24 hours) while tests are being performed.


The wet chemistry test results shall be considered acceptable if the 10B areal density is

determined to be equal to, or greater than, that specified on the fuel tube drawings. The neutron

absorption test shall be considered acceptable if the neutron count determined for each test

specimen is less than or equal to the highest permissible neutron count rate determined from the

BORAL standard, which is based on the 10B areal density specified on the fuel tube drawings.

Any specimen not meeting the acceptance criteria for either test method shall be considered to be

nonconforming material and shall be evaluated within the NAC International QA Program.

Nonconforming material shall be assigned one of the following dispositions: "use-as-is,"

"rework" or "reject." Only material that is determined to meet all applicable conditions of the

license will be accepted.

8.1.9 Transportable Storage Canister

The transportable storage canister is constructed of Type 304L (Yankee-MPC and CY-MPC) or

304/304L (MPC-LACBWR) stainless steel and is fabricated by welding. If circumferential •

8.1-22

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•

NAC-STC SAR

Docket No. 71-9235

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Revision 18B

welds are required to join two shell sections, the seam welds shall not be aligned within 45°

circumferentially. The welded cylinder is closed at the bottom by a circular plate welded to the

shell wall. The top of the cylinder is closed by two field-installed circular plates, welded to the

canister shell wall following fuel loading.

The transportable storage canister is a welded closed component. The canister serves as the

confinement boundary component of the NAC-MPC System during storage of spent fuel in the

vertical concrete cask.

The finished surfaces of all canister welds are visually examined in accordance with ASME Code

Section V, Article 9, to verify that the components are assembled in accordance with the License

Drawings and that the components are free of nicks, gouges, and other damage. The acceptance

criteria for the visually examined welds for the Yankee-MPC and the CY-MPC canisters is in

accordance with ASME Code Section VIII, Division 1, UW-35 and UW-36 and Section III,

Subsection NB, NB-4424 and NB-4427. The acceptance criteria for the visually examined welds

of the MPC-LACBWR canister are in accordance with ASME Code, Section IIl, Subsection NF,

NF-5360 .

The seam and girth welds in the transpo1iable storage canister shell are full-penetration welds

that are radiographic (RT) examined in accordance with ASME Code Section V, Article 2. The

acceptance criteria for the RT-examined welds is that specified in ASME Code Section III,

Subsection NB, Article NB-5320. The canister shell to bottom plate weld is a full-penetration

double-bevel weld with an inside fillet weld that is ultrasonic examined in accordance with

ASME Code Section V, Article 5, with acceptance criteria as specified in ASME Code Section

III, Subsection NB, Article NB-5330. The final surfaces of the seam and girth welds in the

canister and the canister shell to bottom plate weld are also liquid penetrant examined in

accordance with ASME Code Section V, Article 6, with the acceptance criteria being that

specified in ASME Code Section III, Subsection NB, Article NB-5350.

Field installed partial-penetration groove welds attach the shield and structural lids (Yankee

MPC and CY-MPC) or the closure lid (MPC-LACBWR) to the canister shell, and the vent and

the drain port covers to the shield lid (Yankee-MPC and CY-MPC) or the inner and outer vent

and drain port covers to the closure lid (MPC-LACBWR) after the canister is loaded. The

closure ring for the MPC-LACBWR canister is welded to both the canister shell and closure lid

by partial penetration welds. For the Yankee-MPC and CY-MPC canister, the root and final

• surfaces of the shield lid weld are liquid penetrant examined. For the MPC-LACBWR canister,

8.1-23

NAC-STC SAR

Docket No. 71-9235

June 2018

Revision 18B

the closure lid to canister shell weld is progressively liquid penetrant examined at the root, mid

plane and final surfaces. The structural lid to shell weld for the Yankee-MPC and CY-MPC

canisters is progressively liquid penetrant examined at the root, every 3/8-inch weld layer and

final surface. Canister vent and drain port cover welds are liquid penetrant examined at the root

and final surfaces unless the welds are completed in a single pass. Welds completed in a single

pass require only final surface liquid penetrant examination.

All liquid penetrant examinations are completed in accordance with ASME Code, Section V,

Article 6. Acceptance criteria for all liquid penetrant examinations are as specified in ASME

Code Section lII, Division 1, Subsection NB, Article NB-5350.

The Yankee-MPC and CY-MPC canister shield lid welds are helium leakage tested in

accordance with ASME Code Section V, Article 10, Appendix V, using a minimum leak rate test

sensitivity of 1 x 10-7 cm3/sec (helium). The MPC-LACBWR canister closure lid to canister

shell weld is hydrostatically tested following completion of the weld.

The fabricator of the transportable storage canister will establish a written weld inspection plan

in accordance with an approved quality assurance program. The weld inspection plan will

include visual, liquid penetrant, ultrasonic, and radiographic examination. In addition, the weld

inspection plan will identify the welds to be examined, the sequence of the examinations, the

type of examination method to be used, and the criteria for acceptance of the weld in accordance

with the applicable sections of the ASME Code.

8.1.10 HL W Overpack and Basket

The HL W Overpack is constructed of Type 304/304L stainless steel and is fabricated by

welding. If circumferential welds are required to join two shell sections, the seam welds shall

not be aligned within 45° circumferentially. The welded cylinder is closed at the bottom by a

circular plate welded to the shell wall. The top of the cylinder is closed by a field-installed

circular plate, welded to the canister shell wall following HL W canister loading.

The HL W Overpack is a welded closed component.

The finished surfaces of all HL W Overpack welds are visually examined in accordance with

ASME Code Section V, Articles 1 and 9, to verify that the components are assembled in

accordance with the License Drawings and that the components are free of nicks, gouges, and

8.1-24

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Docket No. 71-9235

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Revision 18B

other damage. The acceptance criteria for the visually examined welds for the HL W overpack

are in accordance with ASME Code Section VIII, Division 2, Section 7.5.2.2.

The seam, girth, and shell to bottom plate welds in the HL W overpack shell are full-penetration

welds that are dye penetrant (PT) examined in accordance with ASME Code Section V, Articles

1 and 6. The acceptance criteria for the PT examined welds are those specified in ASME Code

Section VIII, Division 2, Section 7.5.7.2.

Field installed partial-penetration groove welds attach the closure lid to the HL W Overpack shell

after HL W canister loading. The closure lid to canister shell weld is visually examined at the

final surface. Visual examinations are completed in accordance with ASME Code, Section V,

Articles 1 and 9. Acceptance criteria for all visual examinations are as specified in ASME Code

Section VIII, Division 2, Section 7.5.2.2.

The HL W Overpack basket is fabricated from Type 304 stainless steel and is fabricated by

welding. If circumferential welds are required to join two shell sections, the seam welds shall

not be aligned within 45° circumferentially. The five (5) welded HL W cylinder cells are closed

at the bottom by a plate welded to the cell wall where accessible. The top of the HL W cy I ind er

cell is open to allow vertical dry loading of a HLW canister.

The finished surfaces of all HL W overpack basket assembly welds are visually examined in

accordance with ASME Code Section V, Articles 1 and 9 to verify that the components are

assembled in accordance with the License Drawings and that the components are free of nicks,

gouges, and other damage. The acceptance criteria for the visually examined welds for the HL W

overpack basket are in accordance with ASME Code Section III, Subsection NF, NF-5360.

Liquid penetrant (PT) examination will be performed on all HL W Overpack basket welds in

accordance with ASME Code, Section V, Articles 1 and 6. Acceptance criteria for the PT

examined welds shall be in accordance with ASME Code, Section III, Subsection NF, NF-5350.

The fabricator of the HLW Overpack and basket assemblies will establish a written weld

inspection plan in accordance with an approved quality assurance program. The weld inspection

plan will include visual and liquid penetrant examination. In addition, the weld inspection plan

will identify the welds to be examined, the sequence of the examinations, the type of

examination method to be used, and the criteria for acceptance of the weld in accordance with

the applicable sections of the ASME Code .

8.1-25

NAC-STC SAR

Docket No. 71-9235

Figure 8.1-1 Thermal Test Arrangement

I I I

le-,-;=,=-· -1--- - - -l I I

.__~POIIER-=~=Pl'L-Y-------~ 1D HEAlI!IS (2e) I

I I I I L

lC12i\ r "-M '11:11

1Cl4 11)11

-, I I I

I I

I I I I I _J

~~---~---~----~~ ---- - -- - --- - - - -----....::

March 2017

Revision 18

IIIP~T Wll'ERS OR [l;UVN.Dff IN~TI~

TEST SET-UP AND ........_"JEST 111M€

(a) EXTERNAL THERMOCOUPLE LOCATIONS

r=

\ . .. • -=

• , ~

1•.,

DISK PLAN VIEW mrn~

(b) INTERNAL CAVITY AND BASKET THERMOCOUPLE LOCATIONS

8.1-26

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NAC-STC SAR

Docket No. 71-9235

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During periods when the cask is not in use for transport, the periodic leakage rate test need not

be performed on an annual basis, but shall be re-performed prior to returning the cask to service

and use as a transport package.

8.2.2.3 Periodic and Maintenance Leakage Test Acceptance Criteria

8.2.2.3.1 Metallic Seal Testing Acceptance Criteria

The periodic or maintenance leakage testing of the containment boundary closures using metallic

seals consists of a series of leak tests. The acceptance criteria for each metallic seal is a detected

leakage rate of ::S 2 x 10-7 cm3/sec (helium) at a test system sensitivity of< 1 x 10-7 cm3/sec

(helium) or better.

Unacceptable leakage test results shall be cause for rejection of the component tested. Corrective

actions, including repair or replacement of the seals and/or closure component, shall be taken and

documented as appropriate. The leakage test shall be repeated and accepted prior to returning

the cask to service .

8.2.2.3.2 Viton 0-Ring Testing Acceptance Criteria

The periodic or maintenance leakage testing of the containment boundary closures using Viton

0-rings consists of a series of leak tests. The acceptance criteria for the Viton 0-ring seals is a

cumulative (sum of the three individual leakage tests results) detected leakage rate of ::S 9.3 x 10-5

cm3/sec (helium) at a test system sensitivity of< 4.7 x I 0-5 cm3/sec (helium) or better for

standard directly loaded PWR spent fuel assemblies; or leak tight leakage rate of ::S 2.0 x I 0-7

cm3/sec (helium) at a test system sensitivity of< I .0 x 10-7 cm3/sec (helium) or better for directly

loaded HBU PWR spent fuel assemblies.

Unacceptable leakage rate test results shall be cause for rejection of the component tested.

Corrective actions, including repair or replacement of the 0-rings and/or closure component,

shall be taken and documented as appropriate. The leak test shall be repeated and accepted prior

to returning the cask to service.

8.2.3 Subsystems Maintenance

There are no subsystems maintenance requirements on the NAC-STC.

8.2-3

NAC-STC SAR


8.2.4 Valves. Rupture Disks and Gaskets on the Containment Vessel

June2018

Revision l 8B

There are no valves on the NAC-STC packaging providing a containment function. Four quick

disconnects, one each on the vent, drain, inner lid interseal test and interlid ports, are provided

for ease of cask operation.

The quick-disconnect shall be inspected during each cask loading and unloading operation for

proper performance and function. As necessary, the subject quick-disconnect shall be replaced.

The quick-disconnects shall be replaced every two years during transport operations, and

following fuel unloading after extended storage.

There are no rupture disks on the NAC-STC containment vessel.

All 0-rings on the NAC-STC shall be visually inspected for damage during each cask operation.

All metallic 0-rings shall be replaced during each cask loading sequence. Viton 0-rings shall be

replaced annually and as required, based on leak testing results and inspections during

operations. PTFE 0-rings shall be replaced if damage is noted during the visual inspection and

every two years during transport operations.

8.2.5 Shielding

The gamma and neutron shields of the NAC-STC packaging do not degrade with time or usage.

The radiation surveys performed by licensees prior to transport and upon receipt of the loaded

cask provide a continuing validation of the shield effectiveness of the NAC-STC.

8.2.6 Post-Fabrication Thermal Test

Prior to acceptance at the factory, the heat rejection capability of each fabricated NAC-STC

packaging has been confirmed and verified by the thermal test as described in Section 8.1.6.

Prior to each fuel loading in accordance with the operating procedures of Chapter 7, a visual

inspection on the cask including the visual inspection on the radial neutron shield shell (see

Section 8.2.7) will confirm that there is no change of the heat transfer capability of the

packaging. However, a post-fabrication thermal test shall be performed on an operational NAC

STC packaging if, during handling or transport operations, the packaging experiences an adverse

event such as fire, drops or impacts that result in obvious damage to the neutron shield. The

8.2-4

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NAC-STC SAR


June 2018

Revision 18B

post-fabrication thermal test shall be performed in accordance with the fabrication thermal

acceptance test as described in Section 8.1.6 with approved written procedures including defined

acceptance criteria. The packaging shall meet the acceptance criteria of the post-fabrication

thermal test prior to a return to radioactive material transpo11 operations.

8.2.7 Miscellaneous

The transport impact limiters shall be visually inspected prior to each shipment. The limiters

shall be visually inspected for gross damage or cracking to the stainless steel shells in accordance

with approved written procedures and established acceptance criteria. Impact limiters not

meeting the established acceptance criteria shall be rejected until repairs are performed and the

component re-inspected and accepted.

The cask cavity shall be visually inspected prior to each fuel loading. Evidence of gross scoring

of the cavity surface, or build-up of other foreign matter in the cask cavity that could block the

cavity drainage paths shall be cause for rejection of the cask for use until approved maintenance

and/or repair activities have been acceptably completed. The basket assembly for the directly

loaded (uncanistered) or canistered configuration (prior to initial loading) shall be visually

inspected for deformation of the basket disks or tubes. Evidence of damage shall be cause for

rejection of the basket until approved repair activities have been completed, and the basket has

been re-inspected and approved for use.

The radial neutron shield shell shall be visually inspected prior to each fuel loading. Any crack,

gauge, or gross deformation that could indicate damage of the heat transfer fins shall be cause for

rejection of the cask for use until approved maintenance and/or repair activities have been

acceptably completed.

The overall condition of the cask, including the fit and function of all removable components,

shall be visually inspected and documented during each cask use. Components or cask conditions

which are not in compliance with the Certificate of Compliance shall cause the cask to be

rejected for transport use until repairs and/or replacement of the cask or component are

performed, and the component re-inspected and accepted.

The results of the visual inspections, leakage tests, shielding and radiological contamination

surveys; fuel identification information for the package contents; date, time, and location of the

8.2-5

NAC-STC SAR


June 2018

Revision l 8B

cask loading operations; and remarks regarding replaced components shall be included in the

cask loading report for each loaded cask transport. The requirements of the cask loading report

shall be detailed in the NAC-STC Operations Manual.

8.2.8 Maintenance Program Schedule

Table 8.2-1 presents the overall maintenance program schedule for the NAC-STC.

8.2-6

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NAC-STC SAR


June2018

Revision 18B

Table 8.2-1 Maintenance Program Schedule

Task Frequency Cavity Visual Inspection Prior to Fuel Loading Basket Visual Inspection Prior to Fuel Loading 0-ring Visual Inspection Prior to Fuel Loading Outer Lid, Inner Lid and Port Coverplate

Bolt Visual Inspection Prior to installation during each use Radial Neutron Shield Shell Visual Prior to Fuel Loading Inspection Cask Visual

and Proper Function Inspections Prior to each Shipment Lifting and Rotation Trunnion

Visual Inspection Prior to each Shipment Liquid Penetrant Inspection of surfaces Annually during use and accessible welds

Maintenance Periodic Leakage Rate Test For Viton 0-rings, annually or when replaced. of Inner Lid and Port Coverplate 0- For metallic 0-rings, prior to each loaded transport.

rings Preshipment Leakage Rate Test Prior to loaded transport for casks with Viton 0-

rings Transport Impact Limiter Visual Prior to each shipment Inspection Quick-disconnect

Inspection for Proper Function During each Cask Loading/Unloading Operation Quick-disconnect Replacement Every two years during transport operations Metallic 0-ring Replacement Prior to installation for a loaded transport Viton 0-ring Replacement Annually, or more often, based on inspections

during use or leakage test results Inner and Outer Lid Bolt Replacement Every 240 bolting cycles

(Every 20 years at 12 cycles per year) PTFE 0-ring Replacement Every two years during transport operations or as

required by inspection Periodic Leakage Rate Test Performed within 12 months prior to each shipment

for containment boundary Viton 0-rings. No testing needed for out-of-service packaging or for casks provided with containment boundary metallic seals as metallic seals are replaced and maintenance leakage tested during each loading operation.

Post-Fabrication Thermal Test Performed after a cask experiences an adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield. The cask shall pass the pre-fabrication thermal test prior to being used in a subsequent fuel transport.

8.2-7

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NAC-STC SAR


8.4 Cask Body Fabrication ·

8.4.1 General Fabrication Procedures

June 2018

Revision 18B

The NAC-STC cask body is a welded structure of stainless steel plates and forgings. Chemical

Copper lead is poured in place between the inner and outer shells to serve as the main gamma

shielding material. NS-4-FR is poured in place between the neutron shield shell and the outer

shell. NS-4-FR is also form fit between the bottom inner forging and the bottom plate and in the

inner lid. Welding on the NAC-STC shall be performed in accordance with the requirements of

the ASME Code and the American Welding Society (A WS) Structural Welding Code - Steel

(ANSl/A WS D.1-1) as specified on the NAC-STC License Drawings and Section 8.1.1.

The general fabrication procedures for the NAC-STC are summarized, herein, to facilitate an

understanding of the component configurations and the weld locations shown on the license

drawings.

Each of the two inner shell rings (upper and lower) is rolled from Type XM-19 stainless steel

plate and seam welded longitudinally. The outside diameter of each inner shell ring is machined

to the defined transition section dimensions. The minimum length of each Type XM-19 shell ring

shall be in accordance with the License Drawings. The central inner shell sections are each

rolled from Type 304 stainless steel plate and seam welded longitudinally. The number and

length of the individual inner shell sections to be used to obtain the required total inner shell

length is optional. The inner shell sections are girth welded to each other and the inner shell rings

are girth welded on each end of the inner shell. Longitudinal seam welds in adjacent inner shell

sections shall be offset at a minimum of 15 degrees for girth-welded sections.

After initial rough machining and final weld preparation, the top forging and the bottom inner

forging are individually welded to the opposite ends of the inner shell/inner shell ring weldment

to form the cask cavity. The preparation, examination, and acceptance procedures for the welds

are described in Section 8.1.1 and defined on the License Drawings. Following inspection and

acceptance of the welds, the top forging and the outside diameter of the cask cavity weldment are

final machined. Following final machining of both sides of the inner shell, an ultrasonic

thickness test of the inner shell wall of the cask cavity shall be performed to confirm that the wall

thickness of any location on the shell is not Jess than 1.46 inches (37.1 mm). A wall thickness at

any location of less that 1.46 inches (37.1 mm) will be cause for rejection. Rejected areas of the

8.4-1

NAC-STC SAR


June 2018

Revision 18B

shell wall can be repaired by weld overlay using approved written weld overlay procedures.

Following repair, the repaired areas shall be examined in accordance with the original inspection

requirements and acceptance criteria.

Following thickness testing, the cask cavity weldment, which is the NAC-STC primary

containment boundary, shall be hydrostatically tested according to ASME Code, Section III,

Subsection NB-6000, as described in Section 8.1.2.3. The cask cavity weldment is dried, the

primary containment boundary welds are liquid penetrant examined in accordance with ASME

Code, Section V, Article 6, and the welds are accepted in accordance with ASME Code, Section

III, Subsection NB-5350. The cask cavity weldment is then helium leak tested to verify that the

Containment System Fabrication Verification leak rate is satisfied, as described in Section 8.1.3.

Each of the outer shell sections is rolled from Type 304 stainless steel plate and seam welded

longitudinally. The number and length of outer shell sections to be used to achieve the required

total outer shell length is optional. The outer shell sections are girth welded to each other and the

inside diameter of the "outer shell weldment" is final machined. Longitudinal seam welds in

adjacent outer shell sections are offset at a minimum of 15 degrees for gi1ih-welded sections .

The outer shell weldment is welded to the cask cavity weldment at the top forging/outer shell

interface to form the "body weldment." The preparation, examination, and acceptance

procedures for the welds are described in Section 8.1.1 and defined on the License Drawings.

The body weldment is inverted (closure end down) in a pit or other sheltered location m

preparation for lead pouring. A temporary dam extension and supports are welded to the open

end of the outer shell to permit the full length of the lead shell to be poured and to maintain the

outer shell position. "Backing bars" are tack-welded on the inside diameter of the outer shell

overlapping the end of the weld prep and on the top surface of the bottom inner forging

overlapping the outside diameter of the forging (adjacent to the outside diameter of the inner

shell). The backing bars prevent the lead contamination of the welds when the outer shell/bottom

outer forging weld and the bottom outer forging/bottom inner forging weld are performed after

cask body cooldown following the lead pour. Lead pouring preparations, the pour itself, and the

cooldown are performed in accordance with the lead pour requirements and procedures as

described in Section 8.4.2.

Following cooldown, the cask may be moved to a location that is more suitable for the

fabrication activities that are to follow. The temporary dam extension and supports at the open

8.4-2

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NAC-STC SAR


June 2018

Revision l 8B

end of the outer shell are removed and the lead is machined to its final configuration, including

facing off the backing bars to ensure that no lead remains on the weld side of the backing bars.

The bottom outer forging is welded to the outer shell and to the bottom inner forging with the

backing bars preventing lead contamination of the welds. The weld examination and acceptance

criteria are described in Section 8.1.1 and defined on the License Drawings. The NS-4-FR

neutron shield material is installed in the bottom forging of the NAC-STC. The NS-4-FR is

machined to obtain the specified 2-inch thickness and to provide a groove around the outside

diameter. A backing bar is tack-welded on the inside diameter of the bottom outer forging in the

groove in the NS-4-FR and flush with its surface. The bottom plate is positioned and welded to

the bottom outer forging. The weld examination and acceptance criteria are described in Section

8.1.1 and are defined on the License Drawings.

The outside diameter of the outer shell is then machined to the specified final dimensions. If

required to achieve dimensional compliance with the License Drawings, additional localized

machining of the inner shell will be performed. Remachined areas of the inner shell shall be

re-examined by ultrasonic testing to confirm that the minimum thickness of 1.46 inches (37.1

mm) is maintained. Upon completion of final machining and prior to removal from the machine,

the dimensional inspection of the inside diameter and cylindricity of the cavity shall be

performed. Using calibrated M&TE, the inside diameter at 0, 45, 90 and 135 degree radial

locations shall be measured. This measurement shall be repeated at a minimum of 6 axial

locations through the bore of the inner shell. Using calibrated M&TE, a "sweep" of the entire

length of the bore at the same radial locations previously measured and also a "sweep" of the

diameter at the same axial locations will be performed. The combination of these two inspections

will demonstrate the actual diameter and cylindricity of the inner shell bore. Calibrated

inspection equipment and approved written procedures will be used to perform the final

dimensional inspections.

The Type 17-4 PH stainless steel lifting trunnions are welded to the top forging. The Type 17-4

PH stainless steel rotation trunnion recesses are welded to the outer shell at its juncture with the

bottom outer forging. Both the lifting trunnion and rotation trunnion recess weld surfaces are

prepared with a minimum 0.25-inch thick overlay of Inconel. The shear ring and the neutron

shield upper end plate are welded to the top forging. The weld examination and acceptance

criteria are described in Section 8.1.1, and are defined on the License Drawings .

8.4-3

NAC-STC SAR

Docket No. 71-9235

March 2017

Revision 18

The explosively-bonded stainless steel/copper (SS/Cu) heat transfer fins extending through the

neutron shield are welded ( only the stainless steel is welded) to the upper end plate and to the

outer shell. Following liquid penetrant examination of the fin to outer shell welds, the 24 neutron

shield shell plates are prepared for installation and 1/8-inch thick expansion foam is applied to

the interior surface using approved adhesive in accordance with the License Drawings. The

neutron shield shell plates are individually positioned and welded to the stainless steel extended

tip of the SS/Cu fins. These closure welds are then examined and accepted in accordance with

the requirements of the License Drawings. The cask is then placed in the inverted position

(closure end down). Following an installation procedure that has been approved by NAC or by

the material supplier if the material supplier is not NAC, the NS-4-FR neutron shield material is

installed by pouring into each of the 24 regions between the fins in the NAC-STC neutron shield

cavity. After the NS-4-FR has hardened, expansion foam (Section 4.5.3) is installed in the open

end of the neutron shield. The inside and outside diametrical ( curved) surfaces of the expansion

foam are covered by a protective thermal insulation material (Fiberfrax, see Section 4.5.4). The

24 sections of the neutron shield bottom end plate are each positioned and welded to the outer

shell, the fins, the neutron shield shell, and to each other. All of the neutron shield and fin welds

are liquid penetrant examined and accepted in accordance with the License Drawings. The

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neutron shield tank is leak tested using the pneumatic bubble method to verify shell integrity. •

The Type 17-4 PH stainless steel outer lid forging and the Type 304 stainless steel inner lid

forging are machined to the specified final dimensions. The NS-4-FR neutron shield material is

installed in the top of the inner lid following an installation procedure that has been approved by

NAC and by the material supplier. The exposed surface of the NS-4-FR is machined to obtain

the specified 2-inch thickness and the coverplate is welded to the inner lid body. The weld

examination and acceptance are in accordance with the requirements of the License Drawings.

The top surface of the inner lid is then final machined.

The remaining fabrication details (including the installation of the drain line) are then completed.

Following machining of the structural steel support disks and the aluminum heat transfer disks,

the components will be individually inspected for dimensional compliance to the License

Drawings to ensure that each disk meets the stated tolerances. The diameter of each disk is

measured using a calibrated external micrometer. The openings in each disk are inspected using

a calibrated three coordinate measurement machine. The machining center may also be used for

these inspections if previously qualified and calibrated. In the case of the diametral tolerances of

8.4-4 •

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the disks, the inspections are performed at 65 ± 5°F (18 ± 3°C) or else thermal expansion

corrections will be addressed during the inspection process.

The separately fabricated and assembled fuel basket is then inserted into the cask body by

carefully guiding the pre-assembled basket into the cask cavity. The acceptance tests described in

Section 8.1, not previously completed during fabrication, are performed and the completed

NAC-STC is prepared for delivery.

8.4.2 Description of Lead Pour Procedures (Standard Method)

This section describes the general requirements and the standard method procedure that applies

to the pouring of the lead in the annulus formed by the inner and outer shells of the NAC-STC

cask body. The lead annulus provides the primary radial gamma shielding in the cask body and is

subjected to a gamma scan test to verify its shielding integrity. The description that follows

includes the pre-pour preparations, the pouring of the molten lead in the annulus between the

inner and outer shells of the NAC-STC, and the post-pour controlled cooldown of the cask .

8.4.2.1 Preparation for Lead Pour

The following activities must be completed m preparation for pourmg of the lead m the

NAC-STC cask body:

1. Temporary stiffener bars/rings are installed both inside and outside of the body

weldment at intermittent locations along the cask length. The stiffeners support the

inner and outer shells during the lead pour and cooldown in order to maintain the

specified dimensions of the lead annulus. The stiffeners are removed after the

cooldown operation is completed.

2. A minimum of 12 pairs of thermocouples are used to monitor the heating and cooling

cycle of the inner and outer shells. Each pair of thermocouples is positioned at

approximately the same radial and axial location, one on the inside diameter of the

inner shell and one on the outside diameter of the outer shell.

3. Electric heaters are installed in the cask cavity for use in heating the inner shell.

8.4-5

NAC-STC SAR


June 2018

Revision l 8B

4. The body weldment (Section 8.4.1) of the NAC-STC is inverted and suppo1ied in a

stable, vertical position in a "pit" or within a windbreak structure to provide a

basically draft-free operations area.

5. An auxiliary dam extension and supports are welded to the open end of the outer

shell. The extension and supports permit the full length of the lead shell to be poured

in one operation while maintaining the annulus spacing at the open end of the outer

shell.

6. A minimum of 20 gas heating/water cooling rings are installed around the outside of

the body weldment for use in heating, and later in cooling, the outer shell. Gas torches

are provided for heating the outside surface of the bottom inner forging.

7. The body weldment surfaces, especially the lead annulus, are checked for dimensional

accuracy to ensure that the required spacing has been maintained and for cleanliness

to ensure that no foreign materials are present.

•

8. The typical general arrangement of the equipment for the standard lead pour •

operation is shown in Figure 8.4-1.

8.4.2.2 Lead Pour Operations

The requirements and activities that must be completed during the pouring of the lead in the

NAC-STC cask body are:

1. The lead material certification is checked to ensure that it conforms to the

requirements of the American Society of Testing Materials (ASTM) B29, Chemical

Copper Grade - 99.90 percent pure.

2. Approximately 60,000 pounds of lead is placed in appropriate size kettles and melted.

During the lead pouring operations the temperature of the molten lead is maintained

between 650°F (343°C) and 750°F (399°C).

3. At the same time that the lead is being melted, the NAC-STC body weldment is

simultaneously heated using both the electric heaters on the interior and the gas

8.4-6 •

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Revision 18B Docket No. 71-9235, Revision l 7B

8.4.2.3

heating rings on the exterior. The body weldment will be heated in a steady and

uniform manner at a rate not exceeding 125°F/hour (69.4°C/hour). Gas torches are

used to heat the exterior of the bottom inner forging. The surface temperature of the

body weldment is never permitted to exceed 800°F ( 427°C). The temperature of the

entire body weldment is maintained between 640°F (338°C) and 740°F (393°C)

throughout the lead pour operations.

4. The lead pour is initiated immediately after the temperatures of the lead and the body

weldment are stabilized in the ranges previously specified. The actual pouring of the

lead is completed without interruption and in as short a period of time as possible.

During the lead pour the bottom end of the filler-tube is kept below the surface of the

molten lead to preclude the formation of voids in the lead.

5. The lead is poured to a level that is sufficient to ensure that dross removal and

contraction during solidification do not reduce the finished surface below the required

level. A long steel rod inserted into the molten lead annulus is used to ensure that no

solidification has begun anywhere in the volume of molten lead .

Cooldown Following Lead Pour

The procedures and requirements that must be completed during cooldown of the NAC-STC

body weldment following completion of the lead pour are as follows:

1. Cooldown is initiated by turning off the electrical heater (interior) and the gas

heating/water cooling ring (exterior) at the lowest end of the cask (in the as-poured

position). The gas heating/water cooling ring is then used to facilitate and control

cooling by spraying water on the exterior surface of the cask. As cool down proceeds,

the heaters and rings upward along the cask are successively turned off and the

cooling water spray is turned on from each ring.

2. The cooldown process is temperature controlled to maintain approximately uniform

solidification conditions across the thickness and around the circumference of the

annulus .

8.4-7


Revision l 8B Docket No. 71-9235, Revision l 7B

8.4.2.4

3. The cooldown rate is held steady and uniform at a rate not to exceed 125°F/hour

(69.4°C/hour) and the temperature differential between the inside shell and the

outside shell is not allowed to exceed 100°F (55.5°C). Once the inner and outer shell

temperatures have cooled to l 50°F (66°C), it is no longer necessary to control the

cooldown rate.

4. The solidification level in the lead annulus is checked with the aid of a long steel rod.

The maximum difference in the elevation of the solidified lead between the inside

surface of the outer shell and the outside surface of the inner shell is not permitted to

exceed 2 inches (51 mm).

5. Dross is skimmed off the top of the lead while maintaining the molten head

throughout the cooldown process.

Lead Pour Documentation

The following data is included in the Data Package for the Lead Pour Operation:

1. Certificate of Chemical Analysis of the lead.

2. Heating and cooling charts showing elapsed time and temperatures.

3. Location, time and temperature for readings taken with a handheld pyrometer or other

temperature reading device.

4. Difference in solidification elevations when checking at the inside surface of the outer

shell and the outside surface of the inner shell.

8.4.3 Description of Lead Pour Procedures (Alternate Method)

This section describes the general requirements and the alternate method procedure that applies

to the pouring of the lead in the annulus formed by the inner and outer shells of the NAC-STC

cask body. The lead annulus provides the primary radial gamma shielding in the cask body and is

subjected to a gamma scan test to verify its shielding integrity. The description that follows

includes the pre-pour preparations, the pouring of the molten lead in the annulus between the

inner and outer shells of the NAC-STC, and the post-pour controlled cooldown of the cask .

8.4-8

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8.4.3.1 Preparation for Lead Pour

The following activities must be completed 111 preparation for pour111g of the lead 111 the

NAC-STC cask body:

1. Temporary stiffener bars/rings are installed inside the body weldment at intermittent

locations along the cask length. Optional stiffener bars/rings may be installed on the

outside of the body weldment at intermittent locations along the cask length. The

stiffeners support the shells during the lead pour and cooldown in order to maintain

the specified dimensions of the lead annulus. The stiffeners are removed after the

cooldown operation is completed.

2. Pairs of thermocouples are used to monitor the heating and cooling cycle of the inner

and outer shells. Each pair of thermocouples is positioned at approximately the same

radial and axial location, one on the inside diameter of the inner shell and one on the

outside diameter of the outer shell. The exact number of pairs shall be determined

prior to the lead pour .

3. Heaters are installed in the cask cavity for use in heating the inner shell. Typical

heaters include but are not limited to electric, gas, etc.

4. The body weldment (Section 8.4.1) of the NAC-STC is inverted and supported in a

stable, vertical position within a structure that provides a basically draft-free

operational area.

5. An auxi I iary dam extension and supports are welded to the open end of the outer

shell. The extension and supports permit the full length of the lead shell to be poured

in one operation while maintaining the annulus spacing at the open end of the outer

shell.

6. Heating/water cooling systems are installed around the outside of the body weldment

for use in heating, and later in cooling, the outer shell. A heating system is also used

on the outside surface of the bottom inner forging. Heating methods include but are

not limited to electric, gas, etc. Cooling methods include but are not limited to

cooling rings, cooling shells, etc. encompassing the inner and outer shells.

7. The body weldment surfaces, especially the lead annulus, are checked for

dimensional accuracy to ensure that the required spacing has been maintained and for

cleanliness to ensure that no foreign materials are present.

8.4-9

NAC-STC SAR June2018

Revision l 8B Docket No. 71-9235, Revisions l7B and 18A

8.4.3.2

8. The typical general arrangement of the equipment for the alternate method lead pour

operation is shown in Figure 8.4-2.

Lead Pour Operations

The requirements and activities that must be completed during the pouring of the lead in the

NAC-STC cask body are:

1. The lead material certification is checked to ensure that it conforms to the

requirements of the American Society of Testing Materials (ASTM) B29, Chemical

Copper Grade - 99.90 percent pure.

2. Approximately 60,000 pounds of lead is placed in appropriate size kettles and melted.

During the lead pouring operations the temperature of the molten lead is maintained

above a sufficient temperature to conduct the pour but below 790°F ( 421 °C).

3. At the same time that the lead is being melted, 1 the NAC-STC body weldment is

simultaneously heated on both the interior and exterior. The body weldment will be

heated in a steady, uniform, and controlled manner at a rate not exceeding 125°F/hour

(69.4°C/hour). A heating system is also used to heat the exterior of the bottom inner

forging. The surface temperature of the body weldment is never permitted to exceed

800°F (427°C). The temperature of the entire body weldment is maintained between

640°F (338°C) and 740°F (393°C) throughout the lead pour operations.

4. The lead pour is initiated after the temperatures of the lead and the body weldment

are stabilized, as previously described. The actual pouring of the lead is completed

without interruption and in as short a period of time as possible. During the lead pour

the bottom end of the filler-tube is kept below the surface of the molten lead to

preclude the formation of voids in the lead.

5. The lead is poured to a level that is sufficient to ensure that dross removal and

contraction during solidification do not reduce the finished surface below the required

level. A long steel rod inserted into the molten lead annulus is used to ensure that no

solidification has begun anywhere in the volume of molten lead.

8.4-10

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NAC-STC SAR June2018

Revision l 8B Docket No. 71-9235, Revisions l 7B and 18A

8.4.3.3 Cooldown Following Lead Pour

The procedures and requirements that must be completed during cooldown of the NAC-STC

body weldment following completion of the lead pour are as follows:

I. Cooldown is initiated by turning off the interior and exterior heaters at the lowest end

of the cask (in the as-poured position). A water tank cooling system is then used to

facilitate and control cooling. As cask cooling proceeds, effected by raising the

cooling system water level, the heaters upward along the cask are successively turned

off, ensuring the head of lead during cooling remains in the molten state.

2. The cooldown process is temperature controlled to maintain approximately uniform

solidification conditions across the thickness and around the circumference of the

annulus. This is maintained by controlling both the minimum inlet water temperature

and the rate in which the water level is increased. The maximum water level rate of

increase shall be no greater than 23.62 inches (600mm)/hr and the minimum inlet

water temperature shall be greater than 45°F (7°C). Overall cooling rate(s) should be

controlled by lead solidification measurements described below in Step 3 below.

3. The solidification level in the lead annulus is checked with the aid of a long steel rod

in order to verify there is no significant difference in solid surface between the inside

and outside of the annulus.

4. Dross is skimmed off the top of the lead while maintaining the molten head

throughout the cooldown process.

8.4.3.4 Lead Pour Documentation

The following data is included in the Data Package for the Lead Pour Operation:

I. Certificate of Chemical Analysis of the lead.

2. Heating and cooling charts showing elapsed time and temperatures .

8.4-11

NAC-STC SAR


Figure 8.4-1 Typical Arrangement of Lead Pour Equipment (Standard Method)

SCAFFOLD

0 0 0

0 0 0 0

ELECTRICAL HEATERS

AUXIWARY DAM

8.4-12

0 0

HEATING &: COOLING RINGS

STIFFENER RINGS

June 2018

Revision l 8B •

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NAC-STC SAR


Figure 8.4-2 Typical Arrangement of Lead Pour Equipment (Alternate Method)

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lHEFiMOCUPLES ,

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8.4-13

June 2018

Revision 18B

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